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Journal articles on the topic 'Nuclear CFD'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 February 2022

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1

Lee, Gong-Hee. "Review of audit calculation activities on the applicability of CFD software to nuclear safety problems." MATEC Web of Conferences 240 (2018): 05016. http://dx.doi.org/10.1051/matecconf/201824005016.

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The Korea Institute of Nuclear Safety (KINS) is tasked with the technical supporting for approving the safety of specific components or design modifications in nuclear power plant. When a licensee submits analysis to the Nuclear Safety & Security Commission (NSSC) for acceptance, KINS staff reviews this analysis and sometimes conducts independent audit calculations. Though recently licensing applications supported by using Computational Fluid Dynamics (CFD) software are increasing for nuclear safety problems, there is no CFD software which obtains a licensing from the domestic regulatory b

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2

Liu, B., S. He, C. Moulinec, and J. Uribe. "Sub-channel CFD for nuclear fuel bundles." Nuclear Engineering and Design 355 (December 2019): 110318. http://dx.doi.org/10.1016/j.nucengdes.2019.110318.

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3

Boyd, Christopher. "Perspectives on CFD analysis in nuclear reactor regulation." Nuclear Engineering and Design 299 (April 2016): 12–17. http://dx.doi.org/10.1016/j.nucengdes.2015.08.001.

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4

Kutiš, Vladimír, Gabriel Gálik, Jauraj Paulech, Justín Murín, and Vladimír Goga. "CFD Analysis of Dry Cask Nuclear Fuel Storage." Strojnícky časopis - Journal of Mechanical Engineering 69, no. 3 (2019): 75–80. http://dx.doi.org/10.2478/scjme-2019-0032.

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AbstractThe article describes the thermo-hydraulic analysis of a dry cask storage building that is used for the storage of depleted nuclear fuel to determine the viability of a buoyancy driven cooling system. The analysis is performed in the form of steady-state CFD simulations. The resulting temperature distributions are them evaluated based on required operation criteria.

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5

Kutiš, Vladimír, Jakub Jakubec, Juraj Paulech, Gálik Gálik, and Tibor Sedlár. "CFD Analysis of Downcomer of Nuclear Reactor VVER 440." Strojnícky casopis – Journal of Mechanical Engineering 66, no. 2 (2016): 55–62. http://dx.doi.org/10.1515/scjme-2016-0018.

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Abstract The paper is focused on CFD analyses of the coolant flow in the nuclear reactor VVER 440. The goal of the analyses is to investigate the influence of the orifice diameter on the mass flow through individual fuel assemblies in the reactor core. The diameter of orifice can be changed during the operation of a nuclear power plant. Considered boundary conditions in the investigated region of the coolant are based on nominal coolant flow conditions in the nuclear reactor VVER 440.

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6

Wibisono, Andhika Feri, Yacine Addad, and Jeong Ik Lee. "ICONE23-2005 A CFD ASSESSMENT FOR MIXED CONVECTION OF NANOFLUIDS FOR NUCLEAR APPLICATION." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23 (2015): _ICONE23–2—_ICONE23–2. http://dx.doi.org/10.1299/jsmeicone.2015.23._icone23-2_3.

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7

Höhne, T., E. Krepper, and U. Rohde. "Application of CFD Codes in Nuclear Reactor Safety Analysis." Science and Technology of Nuclear Installations 2010 (2010): 1–8. http://dx.doi.org/10.1155/2010/198758.

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Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety (NRS) analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural inte

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8

Mañes, Jorge Pérez, Victor Hugo Sánchez Espinoza, Sergio Chiva Vicent, Michael Böttcher, and Robert Stieglitz. "Validation of NEPTUNE-CFD Two-Phase Flow Models Using Experimental Data." Science and Technology of Nuclear Installations 2014 (2014): 1–19. http://dx.doi.org/10.1155/2014/185950.

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This paper deals with the validation of the two-phase flow models of the CFD code NEPTUNEC-CFD using experimental data provided by the OECD BWR BFBT and PSBT Benchmark. Since the two-phase models of CFD codes are extensively being improved, the validation is a key step for the acceptability of such codes. The validation work is performed in the frame of the European NURISP Project and it was focused on the steady state and transient void fraction tests. The influence of different NEPTUNE-CFD model parameters on the void fraction prediction is investigated and discussed in detail. Due to the co

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9

Morghi, Youssef, Amir Zacarias Mesquita, Jesus Alfonso Puente Angulo, and Ana Rosa Baliza Maia. "SIMULAÇÃO DO ESCOAMENTO EM CONTRACORRENTE ÁGUA/AR EM REATORES NUCLEARES PWR UTILIZANDO CÓDIGO OPENFOAM." HOLOS 6 (December 3, 2018): 92–102. http://dx.doi.org/10.15628/holos.2018.5643.

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A limitação do escoamento em contracorrente, ou inundação, é um fenômeno caracterizado pelo controle que um gás exerce no escoamento de um líquido em sentido contrário. Este efeito tem recebido atenção especial da área nuclear, devido à sua influência no comportamento termofluidodinâmico dos reatores nucleares refrigerados à água pressurizada (Pressurized Water Reactor – PWR), durante um acidente de perda de refrigerante – LOCA (Loss-of-Coolant Accident). A modelagem numérica constitui uma ferramenta fundamental para o desenvolvimento da engenharia nuclear. Este trabalho tem o popósito de demo

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10

Liu, Yan, Qiang Wang, and Peng Fei Zhao. "The Speed Coefficient Method Based Nuclear Main Pump Design and CFD Analysis." Advanced Materials Research 455-456 (January 2012): 302–7. http://dx.doi.org/10.4028/www.scientific.net/amr.455-456.302.

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Nuclear main pumps are one of the most important devices in nuclear plants. Hence hydraulic performances of the nuclear main pump are important. Compared with the traditional speed coefficient method, a revised speed coefficient method is employed to design a nuclear main pump, which is based on the extension chart of the traditional speed coefficient chart. Meanwhile numerical simulations are carried out to examine performances of the designed pump. Results show that performances of the designed pump meet specifications. Clearance leakage causes decrease in head and efficiency of the pump (ab

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11

Smith, Brian L. "ASSESSMENT OF CFD CODES USED IN NUCLEAR REACTOR SAFETY SIMULATIONS." Nuclear Engineering and Technology 42, no. 4 (2010): 339–64. http://dx.doi.org/10.5516/net.2010.42.4.339.

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12

van Driel, Michael R. "Cardioplegia heat exchanger design modelling using computational fluid dynamics." Perfusion 15, no. 6 (2000): 541–48. http://dx.doi.org/10.1177/026765910001500611.

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A new cardioplegia heat exchanger has been developed by Sorin Biomedica. A three-dimensional computer-aided design (CAD) model was optimized using computational fluid dynamics (CFD) modelling. CFD optimization techniques have commonly been applied to velocity flow field analysis, but CFD analysis was also used in this study to predict the heat exchange performance of the design before prototype fabrication. The iterative results of the optimization and the actual heat exchange performance of the final configuration are presented in this paper. Based on the behaviour of this model, both the wat

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13

Jeong, Yeong Shin, Kyung Mo Kim, In Guk Kim, and In Cheol Bang. "CFD Analysis of a Concept of Nuclear Hybrid Heat Pipe with Control Rod." Journal of Fluid Machinery 17, no. 6 (2014): 109–14. http://dx.doi.org/10.5293/kfma.2014.17.6.109.

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14

Dmitriev, S. M., A. E. Khrobostov, A. A. Barinov, and V. G. Glavny. "DEVELOPMENT AND ADAPTATION OF VORTEX REALIZABLE MEASUREMENT SYSTEM FOR BENCHMARK TEST WITH LARGE SCALE MODEL OF NUCLEAR REACTOR." Devices and Methods of Measurements 8, no. 3 (2017): 203–13. http://dx.doi.org/10.21122/2220-9506-2017-8-3-203-213.

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The last decades development of applied calculation methods of nuclear reactor thermal and hydraulic processes are marked by the rapid growth of the High Performance Computing (HPC), which contribute to the active introduction of Computational Fluid Dynamics (CFD). The use of such programs to justify technical and economic parameters and especially the safety of nuclear reactors requires comprehensive verification of mathematical models and CFD programs. The aim of the work was the development and adaptation of a measuring system having the characteristics necessary for its application in the

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15

Gera, B., P. K. Sharma, R. K. Singh, and K. K. Vaze. "CFD Analysis of Passive Autocatalytic Recombiner." Science and Technology of Nuclear Installations 2011 (2011): 1–9. http://dx.doi.org/10.1155/2011/862812.

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In water-cooled nuclear power reactors, significant quantities of hydrogen could be produced following a postulated loss-of-coolant accident (LOCA) along with nonavailability of emergency core cooling system (ECCS). Passive autocatalytic recombiners (PAR) are implemented in the containment of water-cooled power reactors to mitigate the risk of hydrogen combustion. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces below conventional ignition concentration limits and temperature and even in presence of steam. Heat of reaction pr

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16

NARABAYASHI, Tadashi, Michitsugu MORI, and Shuichi OHMORI. "Development of Nuclear Reactor Components by Using X-Ray and CFD." Journal of the Visualization Society of Japan 28-1, no. 2 (2008): 1117. http://dx.doi.org/10.3154/jvs.28.1117.

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17

Mahaffy, John. "DEVELOPMENT OF BEST PRACTICE GUIDELINES FOR CFD IN NUCLEAR REACTOR SAFETY." Nuclear Engineering and Technology 42, no. 4 (2010): 377–81. http://dx.doi.org/10.5516/net.2010.42.4.377.

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18

Ding, Peng, Meilan Chen, Wanai Li, Yulan Liu, and Biao Wang. "CFD simulations in the nuclear containment using the DES turbulence models." Nuclear Engineering and Design 287 (June 2015): 1–10. http://dx.doi.org/10.1016/j.nucengdes.2015.02.021.

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19

Şentürk Lüle, Senem, Uner Colak, Murat Koksal, and Gorkem Kulah. "CFD Simulations of Hydrodynamics of Conical Spouted Bed Nuclear Fuel Coaters." Chemical Vapor Deposition 21, no. 4-5-6 (2015): 122–32. http://dx.doi.org/10.1002/cvde.201407150.

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20

Saifi, Qais, and Otso Cronvall. "Thermal transient finite element computation of a mixing Tee by utilizing CFD results." Rakenteiden Mekaniikka 53, no. 1 (2020): 1–11. http://dx.doi.org/10.23998/rm.76158.

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Thermal distribution and fluctuation in any piping component due to turbulent mixing of flows with different temperatures vary greatly. Usually, computational fluid dynamics (CFD) tools are used for estimation of flows in piping components. Fatigue that results from fluctuating thermal mass flow across the components can be computed by coupling the CFD results with structural mechanics based finite element (FE) results. However, this procedure is laborious and computationally very expensive. A fluid temperature function has been developed in this paper as a function of internal wall coordinate

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21

Jakubec, Jakub, Juraj Paulech, Vladimír Kutiš, and Gabriel Gálik. "Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly." Strojnícky casopis – Journal of Mechanical Engineering 67, no. 1 (2017): 69–76. http://dx.doi.org/10.1515/scjme-2017-0007.

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AbstractThe paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

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22

Krepper, E., R. Rzehak, C. Lifante, and Th Frank. "CFD-modelling of subcooled boiling." Kerntechnik 78, no. 1 (2013): 43–49. http://dx.doi.org/10.3139/124.110311.

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23

Cho, Chungho, Nam-il Tak, Jae-Hyuk Choi, and Yong-Bum Lee. "CFD analysis of the HYPER spallation target." Annals of Nuclear Energy 35, no. 7 (2008): 1256–63. http://dx.doi.org/10.1016/j.anucene.2007.12.011.

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24

Orszulik, Magdalena, Adam Fic, and Tomasz Bury. "CFD modeling of passive autocatalytic recombiners." Nukleonika 60, no. 2 (2015): 347–53. http://dx.doi.org/10.1515/nuka-2015-0050.

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Abstract This study deals with numerical modeling of passive autocatalytic hydrogen recombiners (PARs). Such devices are installed within containments of many nuclear reactors in order to remove hydrogen and convert it to steam. The main purpose of this work is to develop a numerical model of passive autocatalytic recombiner (PAR) using the commercial computational fluid dynamics (CFD) software ANSYS-FLUENT and tuning the model using experimental results. The REKO 3 experiment was used for this purpose. Experiment was made in the Institute for Safety Research and Reactor Technology in Julich (

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25

Liu, Yan, Qiang Wang, and Peng Fei Zhao. "The Speed Coefficient Method Based Nuclear Main Pump Design and CFD Analysis." Advanced Materials Research 455-456 (January 2012): 302–7. http://dx.doi.org/10.4028/scientific5/amr.455-456.302.

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26

Shang, Zhi, and Simon Lo. "CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles." Nuclear Engineering and Design 241, no. 11 (2011): 4427–33. http://dx.doi.org/10.1016/j.nucengdes.2010.09.024.

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27

Wu, C. Y., Y. M. Ferng, C. C. Chieng, and Z. C. Kang. "CFD analysis for full vessel upper plenum in Maanshan Nuclear Power Plant." Nuclear Engineering and Design 253 (December 2012): 285–93. http://dx.doi.org/10.1016/j.nucengdes.2012.08.014.

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28

Yadigaroglu, G. "Computational Fluid Dynamics for nuclear applications: from CFD to multi-scale CMFD." Nuclear Engineering and Design 235, no. 2-4 (2005): 153–64. http://dx.doi.org/10.1016/j.nucengdes.2004.08.044.

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29

Redlinger, Reinhard. "DET3D—A CFD tool for simulating hydrogen combustion in nuclear reactor safety." Nuclear Engineering and Design 238, no. 3 (2008): 610–17. http://dx.doi.org/10.1016/j.nucengdes.2007.02.057.

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30

Shang, Zhi. "CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor." Nuclear Engineering and Design 239, no. 11 (2009): 2562–72. http://dx.doi.org/10.1016/j.nucengdes.2009.07.021.

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31

Rzehak, Roland, and Eckhard Krepper. "CFD for Subcooled Flow Boiling: Parametric Variations." Science and Technology of Nuclear Installations 2013 (2013): 1–22. http://dx.doi.org/10.1155/2013/687494.

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We investigate the present capabilities of CFD for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. Very similar modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant nondimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12) as the working fluid. This facilitated measurements of radial profiles for gas volume fraction, gas velocity, liquid temperature, and bubble size. Robust p

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32

Bestion, D., H. Anglart, D. Caraghiaur, et al. "Review of Available Data for Validation of Nuresim Two-Phase CFD Software Applied to CHF Investigations." Science and Technology of Nuclear Installations 2009 (2009): 1–14. http://dx.doi.org/10.1155/2009/214512.

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The NURESIM Project of the 6th European Framework Program initiated the development of a new-generation common European Standard Software Platform for nuclear reactor simulation. The thermal-hydraulic subproject aims at improving the understanding and the predictive capabilities of the simulation tools for key two-phase flow thermal-hydraulic processes such as the critical heat flux (CHF). As part of a multi-scale analysis of reactor thermal-hydraulics, a two-phase CFD tool is developed to allow zooming on local processes. Current industrial methods for CHF mainly use the sub-channel analysis

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33

Terrier, C., H. Cordier, B. Gaudron, S. Bellet, and W. Hay. "ICONE23-1745 IMPROVING CONFIDENCE IN CFD RESULTS FOR NUCLEAR SAFETY DEMONSTRATION : EXAMPLE OF HETEROGENEOUS INHERENT BORON DILUTION." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23 (2015): _ICONE23–1—_ICONE23–1. http://dx.doi.org/10.1299/jsmeicone.2015.23._icone23-1_368.

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34

Xiong, Jinbiao, Yanhua Yang, and Xu Cheng. "CFD Application to Hydrogen Risk Analysis and PAR Qualification." Science and Technology of Nuclear Installations 2009 (2009): 1–10. http://dx.doi.org/10.1155/2009/213981.

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A three dimensional computation fluid dynamics (CFD) code, GASFLOW, is applied to analyze the hydrogen risk for Qinshan-II nuclear power plant (NPP). In this paper, the effect of spray modes on hydrogen risk in the containment during a large break loss of coolant accident (LBLOCA) is analyzed by selecting three different spray strategies, that is, without spray, with direct spray and with both direct and recirculation spray. A strong effect of spray modes on hydrogen distribution is observed. However, the efficiency of the passive auto-catalytic recombiners (PAR) is not substantially affected

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35

Lindstrøm, Erika Kristina, Jakob Schreiner, Geir Andre Ringstad, Victor Haughton, Per Kristian Eide, and Kent-Andre Mardal. "Comparison of phase-contrast MR and flow simulations for the study of CSF dynamics in the cervical spine." Neuroradiology Journal 31, no. 3 (2018): 292–98. http://dx.doi.org/10.1177/1971400918759812.

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Background Investigators use phase-contrast magnetic resonance (PC-MR) and computational fluid dynamics (CFD) to assess cerebrospinal fluid dynamics. We compared qualitative and quantitative results from the two methods. Methods Four volunteers were imaged with a heavily T2-weighted volume gradient echo scan of the brain and cervical spine at 3T and with PC-MR. Velocities were calculated from PC-MR for each phase in the cardiac cycle. Mean pressure gradients in the PC-MR acquisition through the cardiac cycle were calculated with the Navier-Stokes equations. Volumetric MR images of the brain an

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36

Rzehak, Roland, and Eckhard Krepper. "CFD simulation of DEBORA boiling experiments." Archives of Thermodynamics 33, no. 1 (2012): 107–22. http://dx.doi.org/10.2478/v10173-012-0005-0.

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CFD simulation of DEBORA boiling experimentsIn this work we investigate the present capabilities of computational fluid dynamics for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. This kind of modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant non-dimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12) as the working fluid. This facilitated measurements of radial profiles f

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37

Podila, Krishna, Qi Chen, and Yanfei Rao. "CFD Simulations of Molten Salt Reactor Experiment Core." Nuclear Science and Engineering 193, no. 12 (2019): 1379–93. http://dx.doi.org/10.1080/00295639.2019.1627177.

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38

Ali, Majid, Chang Qi Yan, Zhong Ning Sun, Jian Jun Wang, and Athar Rasool. "CFD Simulation of Throat Pressure in Venturi Scrubber." Applied Mechanics and Materials 170-173 (May 2012): 3630–34. http://dx.doi.org/10.4028/www.scientific.net/amm.170-173.3630.

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In nuclear power plant (NPP), particulate matter and gaseous pollutant release into the environment in severe accidents. To prevent from this disaster, filtered vented containment system (FVCS) containing venturi scrubber is being installed. The present work herein is the CFD simulation of throat pressure in venturi scrubber. A commercial software ANSYS CFX tool has been selected for this research. Euler-Euler regime is used to get the picture of behavior of fluid dynamics inside the venturi scrubber. Gas and liquid interact with each other in throat section of venturi scrubber. The pressure a

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39

Terzuoli, F., M. C. Galassi, D. Mazzini, and F. D'Auria. "CFD Code Validation against Stratified Air-Water Flow Experimental Data." Science and Technology of Nuclear Installations 2008 (2008): 1–7. http://dx.doi.org/10.1155/2008/434212.

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Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validati

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40

Liao, Yixiang, and Dirk Lucas. "Possibilities and Limitations of CFD Simulation for Flashing Flow Scenarios in Nuclear Applications." Energies 10, no. 1 (2017): 139. http://dx.doi.org/10.3390/en10010139.

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41

Debbarma, Ajoy. "CFD Study of Rewetting Nuclear Fuel Rod Bundle by Zig-Zag Jet Impingement." Indian Journal of Science and Technology 8, no. 1 (2015): 1–7. http://dx.doi.org/10.17485/ijst/2016/v9i16/92572.

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42

Debbarma, Ajoy, Krishna Murari Pandey, Deepak Sharma, and Gautam Choubey. "CFD analysis of rewetting vertical nuclear fuel rod by dispersed fluid jet impingement." Perspectives in Science 8 (September 2016): 110–12. http://dx.doi.org/10.1016/j.pisc.2016.04.010.

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43

Malet, J., M. Bessiron, and C. Perrotin. "Modelling of water sump evaporation in a CFD code for nuclear containment studies." Nuclear Engineering and Design 241, no. 5 (2011): 1726–35. http://dx.doi.org/10.1016/j.nucengdes.2011.03.004.

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., Hardik P. Patel. "CFD ANALYSIS OF CALANDRIA BASED NUCLEAR REACTOR: PART-II. PARAMETRIC ANALYSIS OF MODERATOR." International Journal of Research in Engineering and Technology 03, no. 07 (2014): 347–51. http://dx.doi.org/10.15623/ijret.2014.0307059.

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45

Benavides, Julio, Gonzalo Jimenez, Marta Galbán, and Miriam Lloret. "Methodology for thermal analysis of spent nuclear fuel dry cask using CFD codes." Annals of Nuclear Energy 133 (November 2019): 257–74. http://dx.doi.org/10.1016/j.anucene.2019.05.026.

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46

Chen, Chong, Xingjun Wang, Mingjun Wang, et al. "CFD simulation of thermal hydraulic phenomena in enclosed cavity of nuclear power plants." Annals of Nuclear Energy 151 (February 2021): 107953. http://dx.doi.org/10.1016/j.anucene.2020.107953.

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47

Paynter, G. C., C. K. Forester, and E. Tjonneland. "CFD for Engine-Airframe Integration." Journal of Engineering for Gas Turbines and Power 109, no. 2 (1987): 132–41. http://dx.doi.org/10.1115/1.3240015.

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This article provides an assessment of current CFD technology with application to propulsion integration, a definition of research and development needed to extend the technology, and a discussion of numerical error assessment and control. The CFD technology is divided into the elemental areas of the computer system, algorithms, geometry and mesh generation, turbulence modeling, and experimental validation; the current status and major issues in each of these areas are defined. Sources of numerical error are identified and some strategies for determining and controlling these are presented. CF

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48

UCHIDA, Ken, and Shinichiro KAWAMURA. "Error Estimation of Values Obtained from Practical CFD Analysis." Journal of Nuclear Science and Technology 45, no. 7 (2008): 625–33. http://dx.doi.org/10.3327/jnst.45.625.

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UCHIDA, Ken, and Shinichiro KAWAMURA. "Error Estimation of Values Obtained from Practical CFD Analysis." Journal of Nuclear Science and Technology 45, no. 7 (2008): 625–33. http://dx.doi.org/10.1080/18811248.2008.9711461.

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50

Peña-Monferrer, C., J. L. Muñoz-Cobo, and S. Chiva. "CFD Turbulence Study of PWR Spacer-Grids in a Rod Bundle." Science and Technology of Nuclear Installations 2014 (2014): 1–15. http://dx.doi.org/10.1155/2014/635651.

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Nuclear fuel bundles include spacers essentially for mechanical stability and to influence the flow dynamics and heat transfer phenomena along the fuel rods. This work presents the analysis of the turbulence effects of a split-type and swirl-type spacer-grid geometries on single phase in a PWR (pressurized water reactor) rod bundle. Various computational fluid dynamics (CFD) calculations have been performed and the results validated with the experiments of the OECD/NEA-KAERI rod bundle CFD blind benchmark exercise on turbulent mixing in a rod bundle with spacers at the MATiS-H facility. Simula

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