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Relevant bibliographies by topics / Nuclear CFD

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Journal articles

Academic literature on the topic 'Nuclear CFD'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 February 2022

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Journal articles on the topic "Nuclear CFD"

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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