Fracture resistance of HT-9 after irradiation at elevated temperature
Conference
·
· Annu. Book ASTM Stand.; (United States)
OSTI ID:6187086
HT-9, a 12 percent chromium martensitic stainless steel of interest for use in ducts for fast reactors and as a first-wall material in magnetic fusion reactors, was irradiated in the Experimental Breeder Reactor II to obtain the first evaluation of fracture toughness changes produced by high-temperature/high-fluence irradiation in this class of alloy. Charpy V-notch, precracked Charpy, and tension specimens were used to evaluate the tensile and fracture behavior of this steel in the as-received condition, after 5000 h of aging at 427 and 538/degree/C and a 108 deg C shift in the ductile-to-brittle transition after irradiation at 419/degree/and a 108 deg C shift in the ductile-to-brittle transition after irradiation at 419/degree/C. Possible mechanisms for the observed response are discussed.
- Research Organization:
- US Nav Res Lab, Washington, DC, USA
- OSTI ID:
- 6187086
- Report Number(s):
- CONF-800609-
- Conference Information:
- Journal Name: Annu. Book ASTM Stand.; (United States)
- Country of Publication:
- United States
- Language:
- English
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ALLOYS
BREEDER REACTORS
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CORROSION RESISTANT ALLOYS
EBR-2 REACTOR
EPITHERMAL REACTORS
EXPERIMENTAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FRACTURE PROPERTIES
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATERIALS
MECHANICAL PROPERTIES
PHYSICAL RADIATION EFFECTS
POWER REACTORS
RADIATION EFFECTS
REACTOR MATERIALS
REACTORS
RESEARCH AND TEST REACTORS
SODIUM COOLED REACTORS
STAINLESS STEELS
STEELS
TENSILE PROPERTIES
THERMONUCLEAR REACTORS