NUREG/CR--0618
Technical Report
·
OSTI ID:6110313
This report presents the results of Loss-of-Coolant (LOC) Test LOC-11, the first test of the Loss-of-Coolant Accident (LOCA) Test Series conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The primary objective of the test was to evaluate the behavior of pressurized water reactor (PWR) fuel under LOCA conditions similar to those postulated during a simulated double-ended cold leg break in a PWR. Test LOC-11 consisted of four, separately shrouded, fresh fuel rods of PWR design, with initial plenum pressure as a variable. Maximum cladding temperatures of up to 1070/sup 0/K (corresponding to high ductility ..cap alpha..-phase Zircaloy) were sought during Test LOC-11. The fuel rods were exposed to a series of three blowdowns from different power and coolant conditions. The final blowdown resulted in the maximum measured cladding temperature of 1034/sup 0/K. Upon disassembly of the test train the rods were found to be uniformly covered with a dark grey oxide. Posttest results indicated slight cladding circumferential swelling of the pressurized rods and slight collapse of the relatively unpressurized rods. The results are compared with the posttest analyses to aid in understanding the coolant thermal-hydraulic behavior and fuel rod behavior.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 6110313
- Report Number(s):
- TREE-1329
- Country of Publication:
- United States
- Language:
- English
Similar Records
Results of the first nuclear blowdown test on single fuel rods (LOC-11 Series in PBF)
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Conference
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Sat Dec 31 23:00:00 EST 1977
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OSTI ID:6400249
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· Nucl. Technol.; (United States)
·
OSTI ID:5944329
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
DEFORMATION
ENERGY TRANSFER
FLUID MECHANICS
FUEL ELEMENTS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
VERY HIGH TEMPERATURE
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
DEFORMATION
ENERGY TRANSFER
FLUID MECHANICS
FUEL ELEMENTS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
VERY HIGH TEMPERATURE
WATER COOLED REACTORS
WATER MODERATED REACTORS