Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

Journal Article · Wed Nov 01 04:00:00 EDT 2017 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2017.08.037· OSTI ID:1398210

Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

• Research Organization: Pacific Northwest National Laboratory (PNNL), Richland, WA (US)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC05-76RL01830
• OSTI ID: 1398210
• Report Number(s): PNNL-SA-125402; AF5805020
• Journal Information: Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 495; ISSN 0022-3115
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

oxide reduction
pyroprocessing
strontium chloride
tellurite glass