FUNDAMENTAL AND APPLIED RESEARCH AND DEVELOPMENT IN METALLURGY. Progress Report for October 1960

Technical Report · Mon Nov 21 04:00:00 EST 1960

OSTI ID:4086626

In the hot extrusion program only fabrication of the enriched irradiation samples and selection of the irradiation facility remained as objectives. Bend tests are reported on Zircaloy-2-clad U--2 wt.% Zr alloy, aluminum-clad Al--26 wt.% U alloys, and Zircaloy-4-clad uranium honeycomb type fuel elements. Yttrium metal was fabricated with a protective copper cladding at elevated temperatures, without a protective cladding, and at ambient temperstures. All material was taken from the 2-in.-diameter bar prcduced by primary extrusion of the cast ingot. The design of a fuel fabrication plant capable of handling U/ sup 233/ is in progress. The program to determine whether the retained beta phase in uranium-base alloys is stabilized by neutron bombardment and whether the swelling rste of beta -phase uranium is less than that of alpha ursnium was continued. Research into the mechanism of the corrosion of zirconium alloys in hightemperature steam, with emphasis on the role of intermetallic precipitates, was continued. The preparation and thermal analysis of Be --Ce, Be -- Cu, Be -- La, Be -- Mn, Be --Ni--Co, and Be -- Fe --N; alloys are reported. Elevated temperature tensile tests on pure beryllium and a Be--8 at.% Ni alloy are in process. (For preceding period see NMI-2088.) (W.L.H.)

• Research Organization: Nuclear Metals, Inc., Concord, Mass.
• NSA Number: NSA-15-011453
• OSTI ID: 4086626
• Report Number(s): NMI-2090
• Country of Publication: United States
• Language: English

Related Subjects

ALUMINUM
COPPER
CORROSION
ENRICHMENT
EXTRUSION
FABRICATION
FUEL CANS
HIGH TEMPERATURE
INTERMETALLIC COMPOUNDS
IRRADIATION
LANTHANUM ALLOYS
METALLURGY
METALS, CERAMICS, AND OTHER MATERIALS
NEUTRONS
PRECIPITATION
SAMPLING
STABILITY
STEAM
TENSILE PROPERTIES
URANIUM
URANIUM 233
URANIUM ALLOYS
YTTRIUM
ZIRCALOY
ZIRCONIUM ALLOYS