Effect of axial stress on the transient mechanical response of 20%, cold-worked Type 316 stainless-steel cladding
Journal Article
·
· Nucl. Technol.; (United States)
OSTI ID:6431350
To understand the effects of the fuel-cladding mechanical interaction on the failure of 20% cold-worked Type 316 stainless-steel cladding during anticipated nuclear reactor transients, the transient mechanical response of the cladding was investigated using a transient tube burst method at a heating rate of 5.6/sup 0/C/s and axial-to-hoop-stress ratios in the range of 1/2 to 2. The failure temperatures were observed to remain essentially constant for the transient tests at axial-to-hoop-stress ratios between 1/2 and 1, but to decrease with an increase in axial-to-hoop-stress ratios above unity. The uniform diametral strains to failure were observed to decrease monotonically with an increase in axial-to-hoop-stress ratio from 1/2 to 2, and in general, the uniform axial strains to failure were observed to increase with an increase in axial-to-hoop-stress ratio. The fracture of the cladding during thermal transients was found to be strongly affected by the maximum principal stress but not by the effective stress.
- Research Organization:
- Argonne National Lab., IL
- OSTI ID:
- 6431350
- Journal Information:
- Nucl. Technol.; (United States), Journal Name: Nucl. Technol.; (United States) Vol. 42:3; ISSN NUTYB
- Country of Publication:
- United States
- Language:
- English
Similar Records
Controlled biaxial strain-rate testing of 20% cold-worked type 316 stainless steel fast reactor cladding
Controlled biaxial strain-rate testing of 20% cold-worked type 316 stainless steel fast reactor cladding
Measurement of cladding strain during simulated transient tests
Journal Article
·
Sat Oct 01 00:00:00 EDT 1983
· Nucl. Technol.; (United States)
·
OSTI ID:6477263
Controlled biaxial strain-rate testing of 20% cold-worked type 316 stainless steel fast reactor cladding
Journal Article
·
Sat Oct 01 00:00:00 EDT 1983
· Nucl. Technol.; (United States)
·
OSTI ID:6861958
Measurement of cladding strain during simulated transient tests
Conference
·
Fri Jul 20 00:00:00 EDT 1979
·
OSTI ID:5861342
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500 -- Power Reactors
Breeding
36 MATERIALS SCIENCE
360103* -- Metals & Alloys-- Mechanical Properties
ALLOYS
BREEDER REACTORS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
COLD WORKING
CORROSION RESISTANT ALLOYS
CRYSTAL STRUCTURE
EPITHERMAL REACTORS
FABRICATION
FAILURES
FAST REACTORS
FBR TYPE REACTORS
FRACTURES
FUEL CANS
FUEL-CLADDING INTERACTIONS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATERIALS
MATERIALS WORKING
MICROSTRUCTURE
MOLYBDENUM ALLOYS
NICKEL ALLOYS
REACTORS
RUPTURES
STAINLESS STEEL-316
STAINLESS STEELS
STEELS
STRAINS
STRESS ANALYSIS
STRESSES
TEMPERATURE DEPENDENCE
TRANSIENTS
210500 -- Power Reactors
Breeding
36 MATERIALS SCIENCE
360103* -- Metals & Alloys-- Mechanical Properties
ALLOYS
BREEDER REACTORS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
COLD WORKING
CORROSION RESISTANT ALLOYS
CRYSTAL STRUCTURE
EPITHERMAL REACTORS
FABRICATION
FAILURES
FAST REACTORS
FBR TYPE REACTORS
FRACTURES
FUEL CANS
FUEL-CLADDING INTERACTIONS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATERIALS
MATERIALS WORKING
MICROSTRUCTURE
MOLYBDENUM ALLOYS
NICKEL ALLOYS
REACTORS
RUPTURES
STAINLESS STEEL-316
STAINLESS STEELS
STEELS
STRAINS
STRESS ANALYSIS
STRESSES
TEMPERATURE DEPENDENCE
TRANSIENTS