한국전산유체공학회지 (Journal of computational fluids engineering)

한국전산유체공학회 (Korean Society of Computational Fluids Engineering)
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CUPID 코드를 활용한 2×2 봉다발 부수로 유동 해석

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

  • 이재룡 (한국원자력연구원, 열수력안전연구부) ;
  • 박익규 (한국원자력연구원, 열수력안전연구부) ;
  • 김정우 (서울과학기술대학교, 기계시스템디자인공학과)
  • 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)
  • 투고 : 2016.11.07
  • 심사 : 2016.12.21
  • 발행 : 2016.12.31
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초록

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.

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참고문헌

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