An Integrated Process for Recycling of ThO2 Based Mixed Oxide Rejected Nuclear Fuel Pellets

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DOI: 10.4236/wjnst.2017.74024    964 Downloads   2,144 Views  Citations

ABSTRACT

This paper presents a study on the process engineering aspects of relevance to the industrial implementation of ThO2 and (Th, U)O2 mixed oxide (MOX) pellet type fuel manufacturing. The paper in particular focuses on the recycling of thoria based fuel production scrap which is an economically important component in the fuel manufacturing process. The thoria based fuels are envisaged for Advanced Heavy Water Reactor (AHWR) and other reactors important to the Indian Nuclear Power Programme. A process was developed for recycling the chemically clean, off-specification and defective sintered ThO2 and (Th, U)O2 MOX nuclear fuel pellets. ThO2 doesn’t undergo oxidation or reduction and thus, more traditional methods of recycling are impractical. The integrated process was developed by combining three basic approaches of recycling namely mechanical micronisation, air oxidation (for MOX) and microwave dissolution-denitration. A thorough investigation of the influence of several variables as heating method, UO2 content, fluoride and polyvinyl alcohol (PVA) addition during microwave dissolution-denitration was recorded on the product characteristics. The suitability evaluation of the recycled powder for re-fabrication of the fuel was carried out by analyzing the particle size, BET specific surface area, phase using XRD, bulk density and impurities. The physical and chemical properties of recycled powder obtained from the sintered (Th1-y, Uy)O2 (y; 0 - 30 wt%) pellets advocate 100% utilisation for fuel re-fabrication. Recycled ThO2 by integrated process showed distinctly high sinterability compared to standard powder evaluated in terms of surface area and particle size.

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Singh, G. , Khot, P. , Kumar, P. , Baghra, C. , Bhatt, R. and Behere, P. (2017) An Integrated Process for Recycling of ThO2 Based Mixed Oxide Rejected Nuclear Fuel Pellets. World Journal of Nuclear Science and Technology, 7, 309-330. doi: 10.4236/wjnst.2017.74024.

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