Home / Store / Journals / Electronic Articles / Nuclear Technology / Volume 156 / Number 1 / Pages 69-74
Sule Ergun, Jason G. Williams, Lawrence E. Hochreiter, Hergen Wiersema, Marcel Slootman, Marek Stempniewicz
Nuclear Technology / Volume 156 / Number 1 / Pages 69-74
Format:electronic copy (download)
Critical heat flux (CHF) at a natural boiling condition is an important phenomenon for a research reactor having a small-hydraulic-diameter geometry under a large-break loss-of-coolant accident condition. Accurately predicting the CHF under this condition is very important; therefore, the CHF models used in the best-estimate codes must be validated using appropriate experimental data for a given geometry. The present work focuses on validating the CHF calculations and models within the COolant Boiling in Rod Arrays-Two Fluid (COBRA-TF) code by simulating two sets of experiments, which were performed in tubes and annuli with different length-to-diameter ratios. In this work, the cocurrent upflow and downflow correlations developed by Mishima and Nishihara and Holowach et al. and Zuber correlations for the CHF used in COBRA-TF are validated against the experimental data obtained by Monde and Yamaji and Islam et al. Conclusions on the predictive capability of COBRA-TF for the CHF calculations for small-hydraulic-diameter geometry under natural boiling conditions are provided with the description of the correlations and models used.
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