ZIRCONIUM MODERATOR CLADDING IRRADIATED IN THE SODIUM REACTOR EXPERIMENT

Technical Report · Thu Dec 15 04:00:00 EST 1960

OSTI ID:4072032

Gill, J J

Visual examinations, chemical analyses, bend tests, and metallographic examinations were performed on specimens of SRE zircorium cladding material to determine the effects of approximately 4500 hr of reactor operation. Based on the results of this investigation, the SRE zirconium remaining in the core is deemed satisfactory for continued reactor operation under design conditions. During reactor exposure, the general appearance of the zirconium cladding sunfaces exposed to the sodium was unchanged. Several cans ruptured as a result of a swelling of the graphite caused by the entry of sodium into the cans. No grain growth occurred at the normal maximum reactor operating temperature of about 950 deg F. Increases in the hydrogen content, to as much as 1000 ppm, during SRE exposure caused no change in the ductility of the zirconium at normal reactor operating temperature, even though it was very brittle at room temperature. (auth)

• Research Organization: Atomics International. Div. of North American Aviation, Inc., Canoga Park, Calif.
• DOE Contract Number: AT(11-1)-GEN-8
• NSA Number: NSA-15-011706
• OSTI ID: 4072032
• Report Number(s): NAA-SR-5970
• Country of Publication: United States
• Language: English

Related Subjects

BRITTLENESS
CANNING
DUCTILITY
EXPANSION
FAILURES
GRAIN SIZE
GRAPHITE
HYDROGEN
INSPECTION
IRRADIATION
LEAKS
LIQUID METAL COOLANT
MATERIALS TESTING
METALLOGRAPHY
METALS, CERAMICS, AND OTHER MATERIALS
MODERATORS
QUALITATIVE ANALYSIS
QUANTITY RATIO
RADIATION EFFECTS
REACTORS
SODIUM
TENSILE PROPERTIES
ZIRCONIUM