PCI fuel failure analysis: a report on a cooperative program undertaken by Pacific Northwest Laboratory and Chalk River Nuclear Laboratories.
- comps.
Reactor fuel failure data sets in the form of initial power (P/sub i/), final power (P/sub f/), transient increase in power (..delta..P), and burnup (Bu) were obtained for pressurized heavy water reactors (PHWRs), boiling water reactors (BWRs), and pressurized water reactors (PWRs). These data sets were evaluated and used as the basis for developing two predictive fuel failure models, a graphical concept called the PCI-OGRAM, and a nonlinear regression based model called PROFIT. The PCI-OGRAM is an extension of the FUELOGRAM developed by AECL. It is based on a critical threshold concept for stress dependent stress corrosion cracking. The PROFIT model, developed at Pacific Northwest Laboratory, is the result of applying standard statistical regression methods to the available PCI fuel failure data and an analysis of the environmental and strain rate dependent stress-strain properties of the Zircaloy cladding.
- Research Organization:
- Battelle Pacific Northwest Labs., Richland, WA (USA); Atomic Energy of Canada Ltd., Chalk River, Ontario. Chalk River Nuclear Labs.
- DOE Contract Number:
- EY-76-C-06-1830
- OSTI ID:
- 5664136
- Report Number(s):
- NUREG/CR-1163; PNL-2755
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
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210200 -- Power Reactors
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220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
BWR TYPE REACTORS
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FUEL CANS
FUEL ELEMENT FAILURE
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FUEL RODS
FUEL-CLADDING INTERACTIONS
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
MATHEMATICAL MODELS
PHWR TYPE REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
STRESS CORROSION
STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
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210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
210400 -- Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
BWR TYPE REACTORS
CHEMICAL REACTIONS
COMPUTER CALCULATIONS
CORROSION
FUEL CANS
FUEL ELEMENT FAILURE
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
HEAVY WATER COOLED REACTORS
HEAVY WATER MODERATED REACTORS
MATHEMATICAL MODELS
PHWR TYPE REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
STRESS CORROSION
STRESSES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS