NEW TECHNIQUE FOR ESTIMATION OF VOID FRACTION FROM CONDUCTIVITY PROBE SIGNALS IN A SMALL BREAK LOSS-OF-COOLANT ACCIDENT TEST FACILITY.
A scaled test facility of the Babcock & Wilcox raised loop nuclear steam supply system was used to perform small break loss-of-coolant accident testing, thereby, establishing a data base from which plant predictive system codes could be benchmarked. About 250 instruments were used to record the thermal/hydraulic response of the test facility during the transient, of which 36 were conductivity probes. These probes were designed and installed to determine the liquid-steam interface in the facility hot leg, reactor core vessel, and steam generator components. Thus, to date, the primary function of the probe has been limited to liquid level determination during the course of the transient. This study presents a new technique developed to estimate the local average void fraction using the conductivity probe output signal.
author list (cited authors)
Hassan, Y. A., & Rush, G. C.