Numerous issues regarding nuclear plant safety have stimulated experimental and computational efforts associated with the thermal hydraulics of reactor cooling systems. 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. This study presents the data interpretation of the conductivity probe output signals for various tests. It is concluded that the “dry” state (steam) exists when the conductivity probe output voltage falls to the zero value of ∼ 0.05 V, independent of the fluid vapor temperature in which the probe is immersed. The temperature variations may significantly alter the probe output signal when immersed in single-phase water or a two-phase steam and water mixture, due to the change in electrical conductivity of the water with temperature.