INTERPRETATION OF CONDUCTIVITY-SENSITIVE LIQUID-LEVEL TRANSDUCER SIGNALS IN A SMALL BREAK LOSS-OF-COOLANT ACCIDENT TEST FACILITY
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abstract
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.
