Fuel cladding temperature predictions for LOFT LOCE Ll-5. [PWR]
Conference
·
OSTI ID:6607046
Fuel cladding temperature predictions were performed for loss-of-fluid test (LOFT) loss-of-coolant experiment (LOCE) L1-5 using the RELAP4 computer code. Two versions of this code were used, RELAP4/MOD5 and RELAP4/MOD6. Measured thermocouple data have been compared to these predictions to determine the relative accuracy of the two RELAP4 computer code versions and their respective blowdown heat transfer correlations selected to describe the fuel cladding surface temperature response during the subcooled and saturated blowdown phase of the LOCE. LOFT LOCE L1-5 simulated a 200% double-ended offset shear break in the cold leg of a four-loop large pressurized water reactor (LPWR). The initial conditions for the LOCE were: zero power with a nuclear core installed, primary coolant (PC) temperature of 541 K, PC pressure of 15.6 MPa, and PC flow of 176 Kg/sec. The PC pumps were running until the end of the blowdown phase of the LOCE, and cold leg emergency core coolant (ECC) injection initiated at 19 seconds.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 6607046
- Report Number(s):
- CONF-781105-2
- Country of Publication:
- United States
- Language:
- English
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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
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ENERGY TRANSFER
FUEL CANS
HEAT TRANSFER
LOFT REACTOR
LOSS OF COOLANT
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR EXPERIMENTAL FACILITIES
REACTORS
RESEARCH AND TEST REACTORS
SIMULATION
TANK TYPE REACTORS
TEMPERATURE DISTRIBUTION
TEMPERATURE GRADIENTS
TEST REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ENERGY TRANSFER
FUEL CANS
HEAT TRANSFER
LOFT REACTOR
LOSS OF COOLANT
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR EXPERIMENTAL FACILITIES
REACTORS
RESEARCH AND TEST REACTORS
SIMULATION
TANK TYPE REACTORS
TEMPERATURE DISTRIBUTION
TEMPERATURE GRADIENTS
TEST REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS