High temperature in-reactor creep of 20% cold worked FTR cladding

Conference · Thu Dec 01 04:00:00 EST 1977

OSTI ID:6400521

Gilbert, E R; Lovell, A J; Chin, B A

The in-reactor creep behavior of 20% cold worked Type 316 stainless steel tubing was investigated at temperatures ranging from 410 to 720/sup 0/C for neutron fluences up to 5 x 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV) in the experimental breeder reactor (EBR-II). The irradiation environment was found to either retard or enhance creep rates depending upon the irradiation conditions. When compared to thermal control tests on unirradiated specimens, irradiation enhanced creep rates were observed for temperatures below 620/sup 0/C. Above 620/sup 0/C, however, some specimens exhibited less in-reactor creep than thermal creep of unirradiated specimens. This irradiation hardening was observed for strain rates exceeding 5 x 10/sup -4/% per hour. In-reactor and thermal creep rates were found to be similar at lower strain rates. To fully describe the experimentally observed in-reactor creep behavior with an empirical expression, a commutative equation was developed. The development was based as much as possibe on existing equations for thermal creep and irradiation creep. The thermal creep description is comprised of a slow transient, rapid transient, and a tertiary creep term. The tertiary creep term which becomes important for strain rates greater than 5 x 10/sup -4/% per hour has been modified to predict the irradiation hardening phenomena. Similarly, the irradiation-enhanced contributions to in-reactor creep are comprised of a primary irradiation-enhanced creep term, steady-state irradiation enhanced creep term, and a tertiary irradiation enhanced creep term.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA)
• DOE Contract Number: EY-76-C-14-2170
• OSTI ID: 6400521
• Report Number(s): HEDL-SA-1402; CONF-780722-24
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
220600* -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
36 MATERIALS SCIENCE
360106 -- Metals & Alloys-- Radiation Effects
ALLOYS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
COLD WORKING
CORROSION RESISTANT ALLOYS
CREEP
EPITHERMAL REACTORS
FABRICATION
FAST REACTORS
FFTF REACTOR
FUEL CANS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH TEMPERATURE
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
MATERIALS
MATERIALS WORKING
MECHANICAL PROPERTIES
MOLYBDENUM ALLOYS
NICKEL ALLOYS
PHYSICAL RADIATION EFFECTS
RADIATION EFFECTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
STAINLESS STEEL-316
STAINLESS STEELS
STEELS
TEST REACTORS