Parallel-channel effects and long-term cooling during emergency-core cooling in a BWR/4
The effectiveness of the core spray-cooling system (CSCS) and the low-pressure-coolant-injection (LPI) system during a design-basis accident (DBA) in a boiling-water nuclear reactor (BWR) was investigated. This investigation considered a BWR/4 with both intact and broken jet pumps during the short-term and long-term core cooling of a postulated loss-of-coolant accident (LOCA). The experiments were performed in a special parallel channel effects (PCE) test section in which Freon-114 was used to simulate the conditions in a BWR/4. The PCE test section simulated the major regions of a BWR/4, namely: three heated parallel channels; a bypass channel; a bypass-to-channel leakage path; a jet pump; lower and upper plena; a steam separator and standpipe; and a downcomer. Extensive use of glass components in the PCE test section provided visual observation of the thermal-hydraulic phenomena.
- Research Organization:
- Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering
- OSTI ID:
- 5931723
- Report Number(s):
- NUREG/CR-3376; ON: DE83902796
- Country of Publication:
- United States
- Language:
- English
Similar Records
Parallel channel effects and long-term cooling during emergency core cooling in a BWR/4
Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident
An experimental investigation of boiling water reactor parallel channel effects during a postulated Loss of Coolant Accident
Thesis/Dissertation
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:6351031
Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident
Technical Report
·
Tue Nov 30 23:00:00 EST 1982
·
OSTI ID:6634845
An experimental investigation of boiling water reactor parallel channel effects during a postulated Loss of Coolant Accident
Thesis/Dissertation
·
Thu Dec 31 23:00:00 EST 1981
·
OSTI ID:5443823
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BWR TYPE REACTORS
DATA
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL DATA
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
TEMPERATURE GRADIENTS
TEST FACILITIES
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BWR TYPE REACTORS
DATA
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
EXPERIMENTAL DATA
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
TEMPERATURE GRADIENTS
TEST FACILITIES
WATER COOLED REACTORS
WATER MODERATED REACTORS