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Boron carbide is the prime candidate material for neutron absorbers in fast "breeder reactors and the Past Test Reactor. Important data required for design of control rods for thet>e reactors are swelling, gas release, and structural integrity of boron carbide under the expected operating conditions. This paper presents data from irradiations of boron carbide powders in a thermal reactor and powders and pellets in a fast flux reactor. These data are interpreted and discussed in terms of expected performance in a fast reactor. The most important variable in determining irradiation behavior appears to be the mobility of the helium produced relative to the rate it is produced. This precludes extrapolation of data obtained in thermal reactors to predict performance in fast reactors.We conclude that the only accurate way to predict the performance of boron carbide pellets in a fast reactor is to conduct irradiation tests for the same temperatures, times., and burnup levels expected in service.

Six high-density, low-enriched U,Si2~Al dispersion fuel elements have been tested in the Oak Ridge Research Reactor (ORR). The elements were geometrically identical to standard ORR elements. The uranium density in the fuel meat ranged between 4.6 and 5.2 Mg/m. The elements were fabricated by B&W, CERCA, and NUKEM using their normal materials and fabrication practices, with minor modifications necessitated by the new fuel. The U^Si? contained minor amounts of USi, UgSi and/or uranium solid solution. The elements were irradiated in a variety of typical core positions. Three elements were irradiated to approximately normal ORR burnup, and three elements were irradiated twice as long, to average burnups of ~80% of the initially contained U, well above the burnups normally achieved in research and test reactors. Peak burnups of 98% were achieved.

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