ZIRCONIUM HIGHLIGHTS

Technical Report · Thu Jan 01 04:00:00 EST 1959

OSTI ID:4271779

Some aspects of zirconium metallurgical developments are presented. Included is a proposal to initiate a program of irradiation of Zircaloy-2 aimed at producing results which can be used to predict behavior of this alloy in reactor cores. In addition results of a preliminary test which show several of the characteristics of a strain aging material in Zircaloy-2 are presented. Also, the effect of tungsten contamination on corrosion of Zircaloy-2 was examined. Conclusions are given and graphs are presented. Other information is presented concerning an attempt to improve ductility and corrosion resistance of beta-treated Zircaloy-2 and an evaluation of large Zircaloy-2 ingots. (J.R.D.)

• Research Organization: Westinghouse Electric Corp. Bettis Plant, Pittsburgh
• DOE Contract Number: AT(11-1)-GEN-14
• NSA Number: NSA-13-010059
• OSTI ID: 4271779
• Report Number(s): WAPD-ZH-14
• Country of Publication: United States
• Language: English

Related Subjects

CONTAMINATION
CORROSION
DEFORMATION
DIAGRAMS
DUCTILITY
EFFICIENCY
HEAT TREATMENTS
MATERIALS TESTING
METALLURGY AND CERAMICS
PHASE DIAGRAMS
RADIATION EFFECTS
REACTOR CORE
RODS
STABILITY
TUNGSTEN
ZIRCALOY
ZIRCONIUM