Pump two-phase performance program. Volume 3. Transient tests. Final report. [PWR; BWR]
The primary objective of the C-E/EPRI Pump Two-Phase Performance Program was to obtain sufficient steady-state and transient two-phase empirical data to substantiate and ultimately improve the reactor coolant pump analytical model currently used for LOCA analysis. A one-fifth scale pump, which geometrically models a reactor coolant pump, was tested in steady-state runs with single- and two-phase mixtures of water and steam over ranges of operating conditions representative of postulated loss-of-coolant accidents. Transient tests were also run to evaluate the applicability of the steady-state-based calculational models to transient conditions.
- Research Organization:
- Combustion Engineering, Inc., Windsor, CT (USA)
- OSTI ID:
- 6743330
- Report Number(s):
- EPRI-NP-1556(Vol.3)
- Country of Publication:
- United States
- Language:
- English
Similar Records
Pump two-phase performance program. Volume 7. Test facility description. Final report. [PWR; BWR]
Pump two-phase performance program. Volume 8: data processing methods. Final report
Pump two-phase performance program. Volume IV: comparison of steady-state versus transient test results. Final report
Technical Report
·
Mon Sep 01 00:00:00 EDT 1980
·
OSTI ID:6743327
Pump two-phase performance program. Volume 8: data processing methods. Final report
Technical Report
·
Mon Sep 01 00:00:00 EDT 1980
·
OSTI ID:6808062
Pump two-phase performance program. Volume IV: comparison of steady-state versus transient test results. Final report
Technical Report
·
Mon Sep 01 00:00:00 EDT 1980
·
OSTI ID:6771896
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BWR TYPE REACTORS
COOLING SYSTEMS
FLOW RATE
FLUID FLOW
FLUID MECHANICS
HYDRAULICS
LIQUID FLOW
LOSS OF COOLANT
MECHANICS
MOCKUP
PERFORMANCE TESTING
PRESSURE GRADIENTS
PRIMARY COOLANT CIRCUITS
PUMPS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
STRUCTURAL MODELS
TEST FACILITIES
TESTING
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BWR TYPE REACTORS
COOLING SYSTEMS
FLOW RATE
FLUID FLOW
FLUID MECHANICS
HYDRAULICS
LIQUID FLOW
LOSS OF COOLANT
MECHANICS
MOCKUP
PERFORMANCE TESTING
PRESSURE GRADIENTS
PRIMARY COOLANT CIRCUITS
PUMPS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
STRUCTURAL MODELS
TEST FACILITIES
TESTING
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS