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Home / Publications / Journals / Nuclear Technology / Volume 179 / Number 2 / Pages 220-233

Pyrochemical Reprocessing Tests to Collect Uranium Metal from Simulated Spent Oxide Fuel

Yoshiharu Sakamura, Masaaki Akagi

Nuclear Technology / Volume 179 / Number 2 / August 2012 / Pages 220-233

Technical Paper / Reprocessing

A series of pyrochemical reprocessing tests involving pretreatment, electrolytic reduction, and electrorefining processes were conducted using [approximately]100 g of simulated spent oxide fuel. In the pretreatment process, a simulated spent oxide fuel consisting of dense UO2 pellets containing typical fission product elements (strontium, cerium, neodymium, samarium, zirconium, palladium, and molybdenum) was powderized by voloxidation, which corresponds to fuel decladding. Porous oxide pellets were then fabricated from the obtained oxide powder. In the electrolytic reduction process, [approximately]100 g of the porous oxide pellets was loaded in a cathode basket and electrolytic reduction was performed for 7.6 h in a LiCl-Li2O salt at 650°C. The UO2 was reduced to metallic uranium with a reduction yield of 99.2% and a current efficiency of 74%. All the strontium dissolved into the salt. It was verified that the preparation of porous oxide pellets was highly advantageous in improving the rate of oxide reduction. In the subsequent electrorefining process, the reduction product was loaded in an anode basket and electrorefining was performed for 5.8 h in a LiCl-KCl-UCl3 salt at 500°C. Most of the uranium in the reduction product was anodically dissolved in the salt and the refined uranium metal was collected on a stainless steel cathode. Most of the rare earth elements were dissolved in the salt, whereas zirconium, palladium, and molybdenum remained in the anode residue. The concentration of UCl3 in the salt slightly decreased during electrorefining, since U3+ reacted with the oxides contained in the reduction product to form a uranium oxide precipitate.

 
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