Reduction of Solid Uranium Dioxide in Calcium Salts

Nuclear energy has gained crucial importance since it has a minor impact on climate change and greenhouse gas releases; additionally, the other energy sources are insufficient to reach the world's energy needs without nuclear energy. Another sign that the Generation IV International Forum (Kelly, Gen IV International Forum: A decade of progress through international cooperation, 2014) has pointed out is to utilize uranium resources to the maximum and recycle spent nuclear fuel through burn-up in the Generation IV reactor designs, one of which is the molten salt reactor (MSR). Therefore, the MSR can use the spent nuclear fuel as a fresh fuel when the actinides recycle. That reprocessing of spent fuel could be one of the opportunities to contribute to future nuclear energy goals. This study aims to develop a modified pyroprocessing method to prepare molten salt fuels for MSR from spent oxide nuclear fuel that was burned in light water reactors (LWRs). The process diagram illustrated as (1) spent fuel treatment, (2) chopping and voloxidation of spent oxide fuel, (3) oxide reduction of spent fuel, and then depending on the fuel structure and composition for the MSR, it continues by one or two of the following; – electrorefining, – chlorination, and – fluorination. The subject of this study focused on oxide reduction in two categories: chemical reduction and electrochemical reduction. The system designs have been optimized in calcium salts since they have high calcium metal and calcium oxide solubility. The significant results indicated that both methods would substantially reduce the solid uranium dioxide pellet. The chemical reduction will reduce the total solid pellet at 850oC in the composition of 55.73mol%CaCl2-12.37mol%CaF2-26.58mol%Ca-5.32mol%UO2 over 12 hours. The total reduction in the electrochemical test is seen at 850oC during 12 hours with a salt composition of 79mol%CaCl2-17mol%CaF2-4mol%CaO. These oxide reduction mechanisms are convenient ways to reprocess spent oxide fuel from LWRs to utilize in the MSR. Additionally, the reduced fuel is also applicable to using other next-generation reactors. The prospect of this research is the explicit comparison between chemical and electrochemical methods in calcium salts.

Name thesis:degree_name

M.S.

Level thesis:degree_level

masters

Discipline thesis:degree_discipline

Nuclear Engineering

Department dc:contributor.department

Mechanical Engineering

Grantor dc:publisher

Virginia Tech

Year dc:date.issued

2022

Author dc:creator

• Karakaya, Nagihan

Chair dc:contributor.committeechair

• Zhang, Jinsuo

Committee members dc:contributor.committeemember

• Pierson, Mark A.
• Liu, Yang

• Molten salt fuel preparation, Reduction of spent nuclear fuel, Solid uranium oxide reduction

Language dc:language.iso

en_US

Handle dc:identifier.uri

http://hdl.handle.net/10919/111990

OAI identifier oai:identifier

oai:vtechworks.lib.vt.edu:10919/111990