• Title/Summary/Keyword: Tight Fuel Rod Lattice

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Bubbly, Slug, and Annular Two-Phase Flow in Tight-Lattice Subchannels

  • Prasser, Horst-Michael;Bolesch, Christian;Cramer, Kerstin;Ito, Daisuke;Papadopoulos, Petros;Saxena, Abhishek;Zboray, Robert
    • Nuclear Engineering and Technology
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    • v.48 no.4
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    • pp.847-858
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    • 2016
  • An overview is given on the work of the Laboratory of Nuclear Energy Systems at ETH, Zurich (ETHZ) and of the Laboratory of Thermal Hydraulics at Paul Scherrer Institute (PSI), Switzerland on tight-lattice bundles. Two-phase flow in subchannels of a tight triangular lattice was studied experimentally and by computational fluid dynamics simulations. Two adiabatic facilities were used: (1) a vertical channel modeling a pair of neighboring sub-channels; and (2) an arrangement of four subchannels with one subchannel in the center. The first geometry was equipped with two electrical film sensors placed on opposing rod surfaces forming the subchannel gap. They recorded 2D liquid film thickness distributions on a domain of $16{\times}64$ measuring points each, with a time resolution of 10 kHz. In the bubbly and slug flow regime, information on the bubble size, shape, and velocity and the residual liquid film thickness underneath the bubbles were obtained. The second channel was investigated using cold neutron tomography, which allowed the measurement of average liquid film profiles showing the effect of spacer grids with vanes. The results were reproduced by large eddy simulation + volume of fluid. In the outlook, a novel nonadiabatic subchannel experiment is introduced that can be driven to steady-state dryout. A refrigerant is heated by a heavy water circuit, which allows the application of cold neutron tomography.

Large eddy simulation on the turbulent mixing phenomena in 3×3 bare tight lattice rod bundle using spectral element method

  • Ju, Haoran;Wang, Mingjun;Wang, Yingjie;Zhao, Minfu;Tian, Wenxi;Liu, Tiancai;Su, G.H.;Qiu, Suizheng
    • Nuclear Engineering and Technology
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    • v.52 no.9
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    • pp.1945-1954
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    • 2020
  • Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice rod bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W-Co), wall channel and wall channel (W-W), wall channel and center channel (W-C) as well as center channel and center channel (C-C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of rod bundle. The peak frequency values at W-Co channel could have about 40%-50% reduction comparing with the C-C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in rod bundle.

CURRENT STATUS OF THERMAL/HYDRAULIC FEASIBILITY PROJECT FOR REDUCED- MODERATION WATER REACTOR (2) - DEVELOPMENT OF TWO-PHASE FLOW SIMULATION CODE WITH ADVANCED INTERFACE TRACKING METHOD

  • Yoshida, Hiroyuki;Tamai, Hidesada;Ohnuki, Akira;Takase, Kazuyuki;Akimoto, Hajime
    • Nuclear Engineering and Technology
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    • v.38 no.2
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    • pp.119-128
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    • 2006
  • We start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology. As a part of this technology development, we are developing the advanced interface tracking method to improve the conservation of volume of fluid. The present paper describes a part of the development of the twophase flow simulation code TPFIT with the advanced interface tracking method. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results of numerical simulation, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values obtained by the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.