Date of Award

12-2022

Document Type

Thesis

Degree Name

Master of Science (MS)

Department

Materials Science and Engineering

Committee Chair/Advisor

Dr. Ming Tang

Committee Member

Dr. John Ballato

Committee Member

Dr. Kyle Brinkman

Committee Member

Dr. JianhuaTong

Abstract

Electrochemical reprocessing can be used to recycle presently stored nuclear fuel and consists of dissolving that used fuel in molten salt and the waste produced from these processes is a small amount of a high-level salt waste. Vitrification has been selected as the primary means of safely disposing high and low level radioactive waste. This is due to glass’ ability to incorporate many elements within its matrix, and it is chemically durable with the addition of network formers and other glass forming chemicals. With over 90,000 tonnes of nuclear waste in the United States, the avoidance of additional steps required for vitrification would be desired. Salt waste has a low solubility in borosilicate glass, the most prevalent waste form currently, so phosphate glass is one of the waste forms chosen for its immobilization. R2O3 oxides are required to be incorporated into the phosphate glass to replace water soluble -P-O-P- bonds to assist with the poor chemical durability (a crucial property to be controlled for nuclear waste storage). The processing of these glasses occurs at around 1150°C which allow for processing within a current glass melter technology. Characterization techniques were conducted to ensure the homogeneity of the glass and observe the chemical distribution. Leaching and ion irradiation testing was also crucial to this work as the chemical durability and radiation stability for a nuclear waste form are important characteristics to determine its capabilities within an underground long-term storage/repository.

Author ORCID Identifier

0000-0001-8004-5687

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