Swelling of 316 stainless steel and D9 cladding in FFTF (Fast Flux Test Facility)

Conference · Sat Feb 01 00:00:00 EST 1986

OSTI ID:6047164

Makenas, B J

A data base of cladding swelling measurements from postirradiation examinations of FFTF 316 stainless steel clad fuel pins has been developed which is now sufficiently large to compare to previously published cladding swelling data. This comparison of data was motivated by the observation, in the early cycles of FFTF operation, that design swelling correlations had to be adjusted to account for the observed length increases of driver assembly ducts. These original correlations had been developed from nonfueled EBR-II irradiation tests. The recent data indicate that FFTF and EBR-II cladding, from actual fuel pins with similar heats of steel, behave almost identically with respect to fast fluence. The FFTF data further suggest that, above approx.490/sup 0/C, 316 stainless steel (SS) swells in a manner which is essentially independent of irradiation temperature. A comparison of immersion density measurements with the length increases for FFTF driver fuel pins has also shown the 316 SS cladding swelling behavior to be essentially isotropic. D9 cladding (a titanium modified variant of 316 SS) has also been irradiated in FFTF. The data base here is limited, but is large enough to establish the superiority of D9 over 316 SS as an FFTF fuel pin cladding from a swelling point of view.

• Research Organization: Hanford Engineering Development Lab., Richland, WA (USA)
• DOE Contract Number: AC06-76FF02170
• OSTI ID: 6047164
• Report Number(s): HEDL-SA-3389-FP; CONF-860605-45; ON: DE87014365
• Resource Relation: Conference: 13. international symposium on the effects of radiation on materials, Seattle, WA, USA, 23 Jun 1986; Other Information: Portions of this document are illegible in microfiche products
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
FFTF REACTOR
FUEL CANS
ALLOY-D-9
FUEL PINS
PHYSICAL RADIATION EFFECTS
POST-IRRADIATION EXAMINATION
STAINLESS STEEL-316
SWELLING
ALLOYS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
CORROSION RESISTANT ALLOYS
EPITHERMAL REACTORS
FAST REACTORS
FUEL ELEMENTS
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
IRON ALLOYS
IRON BASE ALLOYS
LIQUID METAL COOLED REACTORS
MATERIALS
MOLYBDENUM ALLOYS
NICKEL ALLOYS
NICKEL STEELS
NICKEL-CHROMIUM STEELS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SODIUM COOLED REACTORS
STAINLESS STEELS
STEELS
TEST REACTORS
220600* - Nuclear Reactor Technology- Research
Test & Experimental Reactors
360106 - Metals & Alloys- Radiation Effects
360103 - Metals & Alloys- Mechanical Properties