Analysis of the PBF in-pile large-break LOCA test results with FRAP-T6/BALON-2. [PWR]
Conference
·
OSTI ID:5331370
A series of four, large-break loss-of-coolant accident fuel behavior experiments have been performed in the Power Burst Facility (PBF) at the Idaho Engineering Laboratory. These experiments have been analyzed by using out-of-pile data to understand the phenomenology of zircaloy cladding ballooning and to construct a mechanistic computer code to describe cladding deformation and failure. The code was then used to quantify the influence of rod internal pressure, cladding heatup, and cladding circumferential temperature differences on ballooning and rupture for fresh and irradiated test rods in the PBF. The analysis indicates that the timing and magnitude of cladding circumferential temperature differences are the primary controlling parameters. Both the experimental and the analytical results support the hypothesis that previously irradiated rods exhibit greater cladding strain at failure than do fresh rods because of small local temperature differences within the cladding.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 5331370
- Report Number(s):
- EGG-M-01782; CONF-820604-11; ON: DE82018765
- Country of Publication:
- United States
- Language:
- English
Similar Records
PBF LOCA test LOC-6 fuel-behavior report. [PWR]
Fuel-rod response during the large-break LOCA Test LOC-6. [PWR]
FRAP-T6 uncertainty study of LOCA tests LOFT L2-3 and PBF LLR-3. [PWR]
Technical Report
·
Thu Mar 31 23:00:00 EST 1983
·
OSTI ID:6137957
Fuel-rod response during the large-break LOCA Test LOC-6. [PWR]
Conference
·
Wed Dec 31 23:00:00 EST 1980
·
OSTI ID:5616666
FRAP-T6 uncertainty study of LOCA tests LOFT L2-3 and PBF LLR-3. [PWR]
Conference
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:6807362
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
ALLOYS
COMPUTER CALCULATIONS
DEFORMATION
FUEL CANS
LOSS OF COOLANT
PBF REACTOR
PULSED REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
STRAINS
STRESSES
TANK TYPE REACTORS
THERMAL STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
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
ALLOYS
COMPUTER CALCULATIONS
DEFORMATION
FUEL CANS
LOSS OF COOLANT
PBF REACTOR
PULSED REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
STRAINS
STRESSES
TANK TYPE REACTORS
THERMAL STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS