Transient dispersed flow film boiling of high pressure water in rod bundle geometry

Results are reported from recent experiments investigating dispersed flow film boiling of high pressure water in upflow through a rod bundle. The data, obtained under mildly transient conditions, are compared with correlations currently used to predict heat transfer in these circumstances. Experimental testing was performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility, a highly instrumented, nonnuclear, pressurized-water loop containing 643.66-m (12-ft) long rods (of which 60 are electrically heated). The rods are arranged in a square array typical of 17*17 fuel rod assemblies in late generation PWRs. Data were collected over typical reactor blowdown parameter ranges. Bundle fluid conditions were calculated with a thermal-hydraulic code assuming homogeneous two-phase flow and thermodynamic equilibrium. Comparisons between the experimentally determined heat transfer coefficients and those calculated by the R.S. Dougall-W.M. Rohsenow correlation indicate that the correlation predicted coefficients are as much as a factor of 2.5 greater than those measured. 13 refs