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http://dx.doi.org/10.6112/kscfe.2016.21.4.071

ASSESSMENT OF THE CUPIDCODE APPLICABILITY TO SUBCHANNEL FLOW IN 2×2 ROD BUNDLE  

Lee, J.R. (Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute)
Park, I.K. (Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute)
Kim, J. (Dept. of Mechanical System and Design Engineering, Seoul National Univ. of Science and Technology)
Publication Information
Journal of computational fluids engineering / v.21, no.4, 2016 , pp. 71-77 More about this Journal
Abstract
The CUPID code is a transient, three-dimensional, two-fluid, thermal-hydraulic code designed for a component-scale analysis of nuclear reactor components. The primary objective of this study is to assess the applicability of CUPID to single-phase turbulent flow analyses of $2{\times}2$ rod bundle subchannel. The bulk velocity at the inlet varies from 1.0 m/s up to 2.0 m/s which is equivalent to the fully turbulent flow with the range of Re=12,500 to 25,000. Adiabatic single-phase flow is assumed. The velocity profile at the exit region is quantitatively compared with both experimental measurement and commercial CFD tool. Three different boundary conditions are simulated and quantitatively compared each other. The calculation results of CUPID code shows a good agreement with the experimental data. It is concluded that the CUPID code has capability to reproduce the turbulent flow behavior for the $2{\times}2$ rod bundle geometry.
Keywords
CUPID Code; Rod bundle; Subchannel;
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Times Cited By KSCI : 2  (Citation Analysis)
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