The in-reactor deformation of the PCA alloy
The swelling and in-reactor creep behaviors of the PCA alloy have been determined from the irradiation of pressurized tube specimens in the FFTF reactor. These data have been obtained to a peak neutron fluence corresponding to approximately 80 dpa in the FFTF reactor for irradiation temperatures between 400 and 750/sup 0/C. Diametral measurements performed on the unstressed specimens indicate the possible onset of swelling in the PCA alloy for irradiation temperatures between 400 and 550/sup 0/C and at a neutron fluence corresponding to approx.50 dpa. The creep data suggest a non-linear fluence dependence and linear stress dependence (for hoop stresses less than 100 MPa) which is consistent with the in-reactor creep behavior of many cold worked austenitic stainless steels. These PCA creep data are compared to available 316 SS in-reactor creep data. The in-reactor creep strains for PCA are significantly less than observed in 20% cold worked 316 SS over the temperature ranges and fluences investigated.
- Research Organization:
- Hanford Engineering Development Lab., Richland, WA (USA)
- DOE Contract Number:
- AC06-76FF02170
- OSTI ID:
- 6047157
- Report Number(s):
- HEDL-SA-3422-FP; CONF-860421-79; ON: DE87014382
- Country of Publication:
- United States
- Language:
- English
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ALLOYS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
CORROSION RESISTANT ALLOYS
CREEP
CRYSTAL STRUCTURE
EPITHERMAL REACTORS
FAST REACTORS
FFTF REACTOR
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
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MATERIALS TESTING
MECHANICAL PROPERTIES
MICROSTRUCTURE
MOLYBDENUM ALLOYS
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PHYSICAL RADIATION EFFECTS
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REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
STAINLESS STEEL-316
STAINLESS STEELS
STEELS
SWELLING
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TITANIUM ALLOYS
TUBES