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A Novel Waste Form for Disposal of Spent-Nuclear-Fuel Reprocessing Waste: A Vitrifiable Cement

Mary Lou Dunzik Gougar, Barry E. Scheetz, Darryl D. Siemer

Nuclear Technology

Volume 125 / Number 1 / January 1999 / Pages 93-103


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A cement capable of being hot isostatically pressed into a glass ceramic has been proposed as the waste form for spent-nuclear-fuel reprocessing wastes at the Idaho National Engineering and Environmental Laboratory (INEEL). This intermediate cement, with a composition based on that of common glasses, has been designed and tested. The cement formulations included mixed INEEL wastes, blast furnace slag, reactive silica, and INEEL soil or vermiculite, which were activated with potassium or sodium hydroxide. Following autoclave processing, the cements were characterized. X-ray diffraction analysis revealed three notable crystalline phases: quartz, calcite, and fluorite. Results of compressive strength testing ranged from 1452 to 4163 psi, exceeding the U.S. Nuclear Regulatory Commission (NRC)-suggested standard of >500 psi. From American National Standards Institute/American Nuclear Society 16.1-1986 leach testing, effective diffusivities for Cs were determined to be on the order of 10-11 to 10-10 cm2/s and for Sr were 10-12 cm2/s, which are four orders of magnitude less than diffusivities in some other radwaste materials. Average leach indices (LI) were 9.6 and 11.9 for Cs and Sr, respectively, meeting the NRC standard of LI > 6. The 28-day Materials Characterization Center-1 leach testing resulted in normalized elemental mass losses between 0.63 and 28 g/(m2day) for Cs and between 0.34 and 0.70 g/(m2day) for Sr. Strontium mass losses meet the <1 g/(m2day) industry-accepted standard while Cs losses indicate a process sensitive parameter.

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