Immobilization of Spent Nuclear Fuel in Iron Phosphate Glass
Twenty-four iron phosphate compositions (15 wt% wasteloading) were evaluated to determine their suitability for vitrifying Al-clad, highly enriched uranium, spent nuclear fuel (SNF). In half the compositions melted, 80 wt% of the Al2O3 in the simulated SNF was removed prior to vitrification. All twenty-four compositions formed homogeneous glasses, many at temperatures as low as 1150°C. As little as 2.5 wt% Na2O decreased melt viscosity and increased alumina solubility in those glasses of higher alumina contents (7.2 wt% Al2O3). None of the glasses contained undissolved uranium compounds as has been found in borosilicate glasses containing as little 4.4 wt% UO2. The chemical durability (measured by the product consistency test (PCT)) of the iron phosphate wasteforms is as good as, and in many cases up to 15 times better than the approved reference material (ARM-1) borosilicate glass.
M. G. Mesko and D. E. Day, "Immobilization of Spent Nuclear Fuel in Iron Phosphate Glass," Journal of Nuclear Materials, Elsevier, Jun 1999.
The definitive version is available at http://dx.doi.org/10.1016/S0022-3115(99)00020-3
Materials Science and Engineering
Savannah River Technology Center (U.S.)
Article - Journal
© 1999 Elsevier, All rights reserved.