Evaluation of critical flow for supercritical steam-water. Final report. [PWR]
Experimental measurements of critical flow of sub-cooled water have been carried out on the Winfrith High Pressure Rig over the pressure range 500 to 4500 psi. Four nozzles have been tested to show the relative characteristics of a 0.07'' diameter square-edged orifice, 0.07'' and 0.10'' diameter rounded inlet parallel throat nozzles and a baffled nozzle having the same minimum flow area as a 0.07'' hole. Measurements have also been made of heat transfer in a 0.466'' diameter tube at nozzle inlet conditions. The report describes the experimental equipment, accuracy of measurements and data processing: 283 experiments are presented in tabular and graphical format and compared against other data and predictions. Critical flow rates have been compared against those predicted by the Homogeneous Equilibrium Model, the Extended Henry Fauske and Modified Burnell models and the Bernoulli equation. The heat transfer data have been compared against the correlations of Dittus-Boelter, Jackson-Hall, and Bishop.
- Research Organization:
- UKAEA Atomic Energy Establishment, Winfrith
- OSTI ID:
- 6016145
- Report Number(s):
- EPRI-NP-3086; ON: DE83902250
- Resource Relation:
- Other Information: Portions are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
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ATWS
CRITICAL FLOW
HEAT TRANSFER
NOZZLES
PRESSURIZERS
RELIEF VALVES
PWR TYPE REACTORS
EXPERIMENTAL DATA
REACTOR SAFETY
TEST FACILITIES
THEORETICAL DATA
TWO-PHASE FLOW
ACCIDENTS
CONTROL EQUIPMENT
DATA
ENERGY TRANSFER
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FLOW REGULATORS
FLUID FLOW
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NUMERICAL DATA
REACTOR ACCIDENTS
REACTORS
SAFETY
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220900* - Nuclear Reactor Technology- Reactor Safety
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Nonbreeding
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