AN INVESTIGATION OF DISPERSION HARDENING IN THE URANIUM-URANIUM DIOXIDE SYSTEM-RESEARCH IN PROGRESS

Technical Report · Wed May 18 04:00:00 EDT 1960

OSTI ID:4003741

Arbiter, W

Metallic U is a desirable fuel material for both fast and thermal reactors. However, at low temperatures its anisotropy can result in growth under irradiation with little or no decrease in density, while at temperatures where this effect diminishes (900 to 1000 deg F) other destructive mechanisms are enccuntered. The U increases in volume due to fission gas bubble formation and an inability of the metal to resist the stresses thus imposed. In a study to aileviate these limitations without detriment to neutron economy, a study of dispersion hardening in the U-- U0/sub 2/ system was undertsken. The results of an initial study showed that substantial improvement in both thermal cycling stability and transverse bend strength at elevated temperature are achieved in dispersion-hardened unalloyed U. Further work is in progress in which oxide content and spacing will be controlled, and fabrication procedures studied, in order that optimum structures for the thermal and radiation stability can be achieved ultimately. (auth)

• Research Organization: Nuclear Development Corp. of America, White Plains, N.Y.
• NSA Number: NSA-15-026521
• OSTI ID: 4003741
• Report Number(s): TID-13339
• Country of Publication: United States
• Language: English

Related Subjects

ANISOTROPY
BUBBLES
CERMETS
EXPANSION
FABRICATION
FAILURES
FISSION PRODUCTS
FUEL ELEMENTS
GASES
HEAT TREATMENTS
HIGH TEMPERATURE
LOW TEMPERATURE
MATERIALS TESTING
METALS, CERAMICS, AND OTHER MATERIALS
RADIATION EFFECTS
STABILITY
STRESSES
TEMPERING
TENSILE PROPERTIES
URANIUM
URANIUM DIOXIDE