RELAP4/MOD6 prediction comparisons with LOFT LOCE L2-3 data. [PWR]
Technical Report
·
OSTI ID:6024448
A comparison is presented between RELAP4/MOD6 predicted and experimental measured quantities for Loss-of-Coolant Experiment (LOCE) L2-3 performed in the Loss-of-Fluid Test (LOFT) facility. These data comparisons provide a detailed record for subsequent analysis. Comparisons indicate that the trends in the system hydraulic response were generally well predicted. The core thermal response, in general, was not well predicted due to the code failing to predict the early rewet. It is recommended that future LOCE L2-3 posttest analysis efforts be centered on gaining a better understanding of modeling rewet phenomena with the RELAP4/MOD6 heat transfer surface. Better modeling of the performance of the steam generator secondary side is also needed before modeling small break LOCEs.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 6024448
- Report Number(s):
- LTR-20-104
- Country of Publication:
- United States
- Language:
- English
Similar Records
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Preliminary posttest analysis of LOFT loss-of-coolant experiment L2-2
Conference
·
Sun Dec 31 23:00:00 EST 1978
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OSTI ID:5958995
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Preliminary posttest analysis of LOFT loss-of-coolant experiment L2-2
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Fri Jun 01 00:00:00 EDT 1979
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OSTI ID:6200503
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPARATIVE EVALUATIONS
COMPUTER CALCULATIONS
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
TEMPERATURE GRADIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPARATIVE EVALUATIONS
COMPUTER CALCULATIONS
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
TEMPERATURE GRADIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS