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Roll-clad plate-type fuel elements were developed for use in the Experimental Boiling Water reactor at Argonne National Laboratory. These plates consisted of an uranium-5 wt.% zirconium-1.5 wt.% niobium alloy core clad with Zircaloy-2.
In a continuing program, fabrication characteristics, physical and mechanical properties, and corrosion behavior in air, CO2, NaK, water, and steam were studied for . binary niobium fuel alloys containing 10, 20, 30, 40, 50, and 60 wt.% uranium To evaluate the effects of two major impurities of niobium, oxygen, and zirconium, three niobium base stocks, differing according to the level of these impurities, were used for each alloy. The impurity combinations employed were 600 ppm oxygen and 0.74 wt.% zirconium, 700 ppm oxygen, and 0.17 wt.% zirconium, and 300 ppm oxygen and 0.02 wt.% zirconium, Representative specimens of these alloys retained their hardness up to 900 deg C The 10 and 20 wt.% uraniuin alloys were successfully rorged at 2500 deg F and rolled at 1800 deg F to sheet. Fabrication characteristics of the remaining alloys are under investigation. The 0.2% offset yield strength of the 10 wt.% uranium alloy was 57,200 psi at room temperature and 36,900 psi at 1600 deg F. For the 20 wt.% uranium alloy it was 93,200 psi at room temperature and 71.000 psi at 1600 deg F. The corrosion life of all of the alloys in air at 572 deg F and in CO2 at 600 deg F was superior to that of unalloyed niobium. In 1000- hr exposures to 600 deg F water most of the alloys exhibited corrosion rates only two or three times greater than that of Zircaloy-2. All oi the alloys appear compatible with NaK at 1600 deg F. The impurity combinations employed in the base niobium appeared to have no effect on the corrosion behavior and mechanical properties of the alloys. (auth).
Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.