Abstracts and Available Papers Presented at the
1999 International RERTR Meeting

PROTOTYPIC IRRADIATION TESTING OF HIGH-DENSITY U-Mo ALLOY DISPERSION FUELS

S. L. Hayes, M. K. Meyer, C. R. Clark and J. R. Stuart
Argonne National Laboratory
Engineering Division
P. O. Box 2528
 Idaho Falls, ID 83403-2528 USA

I. G. Prokofiev and T. C. Wiencek
 Argonne National Laboratory
Energy Technology Division
9700 Cass Avenue
 Argonne, IL 60439-4803 USA

ABSTRACT

An irradiation test vehicle has been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho.  Irradiation of this experiment began in October 1999.  This irradiation test is designed to obtain irradiation performance data on a series of high-density U-Mo alloy dispersion fuels at prototypic reactor conditions.  The irradiation experiment contains 47 miniature fuel plates.  The U-Mo alloy fuel compositions include:  U-10Mo, U-8Mo, U-7Mo, U-6Mo, U-6Mo-1.7Os and U-6.1Mo-0.9Ru, all fabricated at densities >8 g-U/cm3.  U3Si2 fuel plates have also been included in the test matrix at a density of 6 g-U/cm3.  A variety of fabrication techniques were used to produce the fuel powders, including machining, atomization, arc-sputtering and comminution.  Fuel alloys are being tested in both as-fabricated and heat treated conditions.  The U-6Mo alloy is being tested in both aluminum and magnesium matrices, while all others are being tested in an aluminum matrix only.  The peak fuel temperatures are >200°C.  The experiment will be discharged at peak fuel burnups of greater than 30 at.% U235.  Of particular interest are the extent of reaction of the fuel and matrix phases and the fission gas retention/swelling characteristics under conditions of fissile uranium densities and fuel temperatures prototypic of today’s highest enrichment, highest power research reactors.  This paper presents the design of the irradiation test and the irradiation conditions.