Tensile property correlations for highly irradiated 20% C. W. type 316 stainless steel. [Irradiation at 370 to 816/sup 0/C at fast neutron fluence of 8. 4 x 10/sup 22/ n/cm/sup 2/ (E > 0. 1 MeV)]
Experiments on developmental FFTF cladding (20% C.W. Type 316 stainless steel) extended the data base to a fast neutron fluence of 8.4 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV). The specimens were irradiated at temperatures ranging from 371/sup 0/ to 816/sup 0/C, although peak fluence levels were attained on specimens irradiated near 371/sup 0/C and 649/sup 0/C. Tension tests were performed at 232/sup 0/C, near the irradiation temperature, and in some cases, above the irradiation temperature. Test specimen strain rates ranged from 4 x 10/sup -5//s to 4 x 10/sup -2//s. Data generated on cladding irradiated near 371/sup 0/C established that the low temperature strength and ductility are fluence independent beyond about 5 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV). Strength behavior of the irradiated cladding at 538/sup 0/, 593/sup 0/, and 649/sup 0/C is essentially the same as exhibited by thermally aged developmental cladding at the same temperatures and times. Up to a fluence of approximately 5 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV), the 538/sup 0/C ductility values remain relatively fluence independent after an initial decrease. Higher temperature (593/sup 0/C and 649/sup 0/C) ductilities decrease continually with increasing fluence. Tensile parameter correlations were developed for the prediction of irradiation effects on the tensile properties of 20% C.W. Type 316 stainless steel. These correlations are based on unirradiated tensile property correlations developed using Hart's equation of state. The condition of plastic deformation of materials such as 316 stainless steel, can be characterized by a structure parameter (sigma*) which describes the material's ''hardness.'' Irradiation effects can be incorporated into this formulation by parameterizing the changes in sigma* with irradiation temperature and fluence. Correlations provide a description of strength and ductility over the temperature range of 371/sup 0/C to 871/sup 0/C and strain rates of 10/sup -5/ to 10/sup 1//s.
- Research Organization:
- Hanford Engineering Development Lab., Richland, WA (United States)
- DOE Contract Number:
- EY-76-C-14-2170
- OSTI ID:
- 6551330
- Report Number(s):
- HEDL-SA-1462; CONF-780722-3
- Resource Relation:
- Conference: 9. symposium on effects of radiation in structural materials, Richland, WA, USA, 10 Jul 1978
- Country of Publication:
- United States
- Language:
- English
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Mechanical properties of highly irradiated 20 percent cold-worked Type 316 stainless steel
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FFTF REACTOR
FUEL CANS
STAINLESS STEEL-316
PHYSICAL RADIATION EFFECTS
TENSILE PROPERTIES
COLD WORKING
DUCTILITY
ELONGATION
FUEL ELEMENTS
NEUTRON FLUENCE
PLASTICITY
STRAIN RATE
TEMPERATURE DEPENDENCE
YIELD STRENGTH
ALLOYS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
CORROSION RESISTANT ALLOYS
EPITHERMAL REACTORS
FABRICATION
FAST REACTORS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
MATERIALS
MATERIALS WORKING
MECHANICAL PROPERTIES
MOLYBDENUM ALLOYS
NICKEL ALLOYS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
STAINLESS STEELS
STEELS
TEST REACTORS
360106* - Metals & Alloys- Radiation Effects
360103 - Metals & Alloys- Mechanical Properties
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors