TRANSIENT VOID FRACTION PREDICTIONS AT ELEVATED PRESSURES IN ONCE-THROUGH INTEGRAL SYSTEM
A scaled test facility of the Babcock & Wilcox Company raised-loop nuclear steam supply system was used to perform small break loss-of-coolant accident testing. About 250 instruments were used to record the thermal-hydrauliÑ response of the facility during various tests, of which 36 were conductivity probes. These probes were designed and installed to determine the liquid/steam interface in facility components.
This study presents the local average void fractions at the conductivity probe locations using a simple technique. The technique utilizes calibration as a function of temperature for a liquid immersed probe. Favorable agreement between conductivity probe void fractions of test 2202AA and the collapsed liquid levels inferred from the hot leg differential pressure transmitter was obtained. Furthermore, a comparison between the void fraction calculated using RELAP5/M0D2 best-estimate thermal-hydraulic computer code and the calculated void fraction using the new technique was performed. A fair agreement was obtained.