THERMAL AND HYDRAULIC EXPERIMENTS FOR PRESSURIZED WATER REACTORS

Technical Report · Fri Oct 31 04:00:00 EST 1958

OSTI ID:4305470

Esselman, W H; Mandil, I H; Green, S J; Ostergaard, P C; Fredrickson, R A

The results of as analytical and experimental program to obtain knowledge concerning thermal design unknowns for pressurized water reactors are described. Experimental techniques used are described. The outof-pile tests simulated nuclear heating by using elecrically heated test sections. The in-pile experiments were performed with specially instrumented and loaded fuel elements installed in a reactor to measure appropriate thermal and hydraulic conditions. Results on the following subjects are included: experiments on the surface effects of nucleate boiling; pressure drop tests; heat transfer burnout tests; transient loss-of-coolantflow experiments with an operating reactor; and reactor performance during operation at limiting thermal and hydraulic conditions. (M.H.R.)

• Research Organization: Westinghouse Electric Corp. Atomic Power Div., Pittsburgh; Division of Reactor Development, AEC
• NSA Number: NSA-12-015005
• OSTI ID: 4305470
• Report Number(s): A/CONF.15/P/453
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

ANALOG SYSTEMS
BURNOUT
CONFERENCE
COOLANT LOOPS
FAILURES
FLUID FLOW
FUEL ELEMENTS
HEAT TRANSFER
HEATING
HIGH TEMPERATURE
INSTRUMENTS
LEAKS
MATERIALS TESTING
MEASURED VALUES
MELTING
NUCLEATE BOILING
PHYSICS
PLANNING
PRESSURE
REACTOR CORE
REACTORS
SURFACES
THERMAL STRESSES
THERMODYNAMICS
WATER COOLANT
ZONES