Measurement of cladding strain during simulated transient tests
Conference
·
OSTI ID:5861342
A diametral extensometer was developed and employed during temperature ramp tests with the Fuel Cladding Transient Tester (FCTT). Plastic strain measurements were performed using unirradiated 20% cold-worked AISI 316 stainless steel tubing ramped at 5.6 and 111/sup 0/C/s with internal pressures from 3.4 to 93.1 MPa. Results demonstrated that plastic deformation can occur at stresses well below the conventional 0.2% yield strength and that most deformation in such tests occurs in the final 50/sup 0/C before failure. Postirradiation tests were performed on fuel pin cladding irradiated to 5.8 x 10/sup 22/ n/cm (E > 0.1 MeV) with irradiation temperatures to 540/sup 0/C. The tests showed that, for test pressures of 17.2 MPa or less, the stress-strain behavior was unchanged from unirradiated material behavior although the strains at failure were greatly decreased.
- Research Organization:
- Hanford Engineering Development Lab., Richland, WA (USA)
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- EY-76-C-14-2170
- OSTI ID:
- 5861342
- Report Number(s):
- HEDL-SA-1731-FP; CONF-790816-73
- 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
Effects of multiple transients on fast reactor fuel pin cladding mechanical properties
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
Effects of multiple transients on fast reactor fuel pin cladding mechanical properties
Conference
·
Sat Dec 31 23:00:00 EST 1977
·
OSTI ID:6349132
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
BREEDER REACTORS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
CORROSION RESISTANT ALLOYS
DEFORMATION
EPITHERMAL REACTORS
EXTENSOMETERS
FAST REACTORS
FBR TYPE REACTORS
FUEL CANS
FUEL ELEMENT FAILURE
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH TEMPERATURE
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATERIALS
MOLYBDENUM ALLOYS
NICKEL ALLOYS
REACTOR ACCIDENTS
REACTORS
STAINLESS STEEL-316
STAINLESS STEELS
STEELS
STRAINS
TRANSIENTS
VERY HIGH TEMPERATURE
210500 -- Power Reactors
Breeding
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ALLOYS
BREEDER REACTORS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
CORROSION RESISTANT ALLOYS
DEFORMATION
EPITHERMAL REACTORS
EXTENSOMETERS
FAST REACTORS
FBR TYPE REACTORS
FUEL CANS
FUEL ELEMENT FAILURE
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH TEMPERATURE
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATERIALS
MOLYBDENUM ALLOYS
NICKEL ALLOYS
REACTOR ACCIDENTS
REACTORS
STAINLESS STEEL-316
STAINLESS STEELS
STEELS
STRAINS
TRANSIENTS
VERY HIGH TEMPERATURE