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 2×2 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 2x2 rod bundle geometry.
| Translated title of the contribution | ASSESSMENT OF THE CUPID CODE APPLICABILITY TO SUBCHANNEL FLOW IN 2×2 ROD BUNDLE |
|---|---|
| Original language | Korean |
| Pages (from-to) | 71-77 |
| Number of pages | 7 |
| Journal | 한국전산유체공학회지 |
| Volume | 21 |
| Issue number | 4 |
| DOIs | |
| State | Published - 2016 |