Validation of a drift-flux model used in the CATHARE code for rod bundle geometry at low pressure and low liquid flow conditions
Résumé
Nuclear reactor safety is regularly demonstrated with system scale computation codes. Twophase situations are usually modelled using either a two-fluid 6-equation model or a drift-flux model with only a mixture momentum equation and an algebraic relation for predicting the slip between phases. Several codes, like the reference French nuclear safety code CATHARE, derive the interfacial friction model of the phase momentum equations from a drift flux model for the bubbly-slug-churn flow regimes.
Currently, the Bestion’s drift-flux model developed in 1990 is used in the CATHARE code interfacial friction for rod bundle geometry. It has replaced the first Bestion’s drift-flux model developed in 1985. The drift velocity assumes that large distorted bubbles crossed by several rods characterize the flow in rod bundle geometry. The model and its parameters were validated on a large database but some doubt exists on the validity at very low pressure such as atmospheric pressure.
Experimental tests were performed at Purdue University (West Lafayette, USA) with a geometry representative of a boiling water reactor core, and provided new in air-water data at atmospheric pressure and low liquid flow conditions. The CATHARE model is compared against these data. Despite experimental conditions are out of the current validation range, good agreement is observed.
Other recent experimental studies with advanced measurement techniques such as transient tomography and wire mesh sensors seem to corroborate the existence of large bubbles occupying several sub-channels. Some potential improvements of the drift-flux model are discussed like the friction coefficient between the large bubbles and the liquid film along the rods.