A 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 and 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-stream 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. The technique uses calibration data as a function of temperature for a liquid immersed probe and the singular value when immersed in steam to estimate the void fraction in the presence of a two-phase steam and water mixture. Favorable agreement between conductivity probe void fraction and that inferred from a differential pressure transmitter was obtained.
- OSTI ID:
- 5537876
- Report Number(s):
- CONF-851125-; TRN: 86-024175
- Resource Relation:
- Conference: American Society of Mechanical Engineers winter annual meeting, Miami, FL, USA, 17 Nov 1985
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
42 ENGINEERING
LOSS OF COOLANT
HYDRAULICS
THERMAL ANALYSIS
TWO-PHASE FLOW
VOID FRACTION
CALCULATION METHODS
CALIBRATION
DATA BASE MANAGEMENT
INTERFACES
LEADING ABSTRACT
MEETINGS
PRESSURE GRADIENTS
PROGRAMMING
REACTORS
STEAM
STEAM GENERATORS
TEMPERATURE DEPENDENCE
TEST FACILITIES
THERMAL CONDUCTIVITY
TRANSIENTS
WATER
ABSTRACTS
ACCIDENTS
BOILERS
DOCUMENT TYPES
FLUID FLOW
FLUID MECHANICS
HYDROGEN COMPOUNDS
MANAGEMENT
MECHANICS
OXYGEN COMPOUNDS
PHYSICAL PROPERTIES
REACTOR ACCIDENTS
THERMODYNAMIC PROPERTIES
VAPOR GENERATORS
220900* - Nuclear Reactor Technology- Reactor Safety
420400 - Engineering- Heat Transfer & Fluid Flow