Recent R&D Studies Related to Coprocessing of Spent Nuclear Fuel Using N,N-Dihexyloctanamide

Abstract

The PUREX process has undergone several modifications to address the issues of high burn up, fewer solvent extraction cycles, and reduced waste arisings. Advanced fuel cycle scenarios have led to a renewed international interest in the development of separation schemes for co-recovering U/Pu from spent fuels. Completely incinerable N,N-dihexyloctanamide (DHOA) has been identified as a promising candidate for the reprocessing of spent fuels. Batch extraction studies were carried out to evaluate DHOA and TBP for the coprocessing (co-extraction and co-stripping) of U and Pu from spent fuel under varying concentrations of nitric acid and of uranium as well as under simulated pressurized heavy water reactor spent fuel feed conditions. At 50 g/L U in 4 M HNO₃, D_Pu values for 1.1 M DHOA and 1.1 M TBP solutions in n-dodecane were 7.9 and 3.8, respectively. In contrast, significantly lower D_Pu value at 0.5 M HNO₃ (4 × 10⁻³) for DHOA as compared to TBP (4 × 10⁻²) suggested that it was a better choice for coprocessing of spent nuclear fuel. This behavior was attributed to the change in stoichiometry of extracted species at lower acidity vis-a-vis the higher acidity. These studies suggest that plutonium fraction can be enriched with respect to uranium contamination in the product stream. DHOA displays better extraction behavior of plutonium and stripping behavior of uranium under simulated feed conditions. DHOA appears distinctly better than TBP with respect to fission product/structural material decontamination of U/Pu.

Keywords:

• Coprocessing
• dihexyloctanamide
• plutonium
• PUREX
• TBP
• uranium