Loss-of-Coolant Accident Test Series Test LOC 5 Experiment Predictions
The Loss of Coolant Accident (LOCA) Test Series being conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory has been designed to provide data for the development and the assessment of fuel behavior computer codes used to predict the response of a pressurized light water reactor (PWR) during a hypothetical break in the cold-leg inlet or hot-leg outlet. This report presents the experiment predictions for the four-rod LOCA test, LOC-5. An analysis was performed to predict the test fuel rod and system behavior during a typical LOC test. Reactor physics calculations were performed with the RAFFLE code to determine the relationship between test fuel rod powers and the PBF reactor power during both the steady state operation and during the blowdown. Calculation of the system thermal-hydraulic response during blowdown, made with the RELAP4 computer code, provided the coolant and heat transfer boundary conditions for the fuel behavior calculations. Cladding and fuel rod dimensions for the rods previously irradiated in the Saxton reactor were determined with the FRAP-S code. Finally, the rod thermal and mechanical behavior during the blowdown transient were determined with the FRAP-T5 code.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 5958148
- Report Number(s):
- TFBP-TR-332; INL/HST-23-75829
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
Fuel element failure
Fuel elements
Heat transfer
Nuclear Reactor Technology- Reactor Safety
PWR Type Reactors
Reactor accidents
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210100 -- Power Reactors
Nonbreeding
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220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ENERGY TRANSFER
FUEL ELEMENT FAILURE
FUEL ELEMENTS
HEAT TRANSFER
LOSS OF COOLANT
Loss of Coolant
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SWELLING
WATER COOLED REACTORS
WATER MODERATED REACTORS
Fuel elements
Heat transfer
Nuclear Reactor Technology- Reactor Safety
PWR Type Reactors
Reactor accidents
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
ENERGY TRANSFER
FUEL ELEMENT FAILURE
FUEL ELEMENTS
HEAT TRANSFER
LOSS OF COOLANT
Loss of Coolant
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SWELLING
WATER COOLED REACTORS
WATER MODERATED REACTORS