Loss-of-coolant accident test series test loc 3 experiment predictions. [PWR]
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 light water reactors during a hypothetical break in the cold-leg inlet or hot-leg outlet of a pressurized water reactor (PWR). This report presents the experiment predictions for the four-rod LOCA test, LOC-3. Test LOC-3 will be performed using Saxton design fuel rods. Two of the fuel rods are unirradiated and two rods are irradiated to about 16,000 MWd/t, and each rod is surrounded by an individual coolant flow shroud. One of each set of the irradiated and the unirradiated rods will be backfilled with helium to a pressure typical of beginning-of-life PWR fuel rods, and the other rods will be backfilled to a pressure typical of fuel rods at the end-of-operational life.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 6207515
- Report Number(s):
- TFBP-TR-314; TRN: 79-015746
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL RODS
PERFORMANCE TESTING
LOSS OF COOLANT
PRESSURE GRADIENTS
TEMPERATURE GRADIENTS
PWR TYPE REACTORS
PBF REACTOR
REACTOR SAFETY
TEST FACILITIES
ACCIDENTS
FUEL ELEMENTS
PULSED REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
TANK TYPE REACTORS
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled