Abstract
The alkaline dissolution process employed in the routine production of the important medical isotope 99Mo generates an inhomogeneous residue that contains enriched uranium contaminated with fission, activation and transuranium products. Development of a process to recover and purify the uranium will be beneficial in unlocking its commercial value and alleviating storage space constraints for this highly radioactive residue. In this study an ammonium carbonate-based process was established that can be used to recover uranium with high yields and purify it to such an extent that it can be removed from a shielded cell for final purification. As a unique application of this dissolution medium, this is the first attempt at uranium recovery from a commercial 99Mo production process. The residue was characterized in terms of uranium content (70% in unirradiated simulant residue, but only 47% in real irradiated residue), uranium oxidation state (U4+:U6+ ratio of 11:89) and chemical speciation (U6+ present as a mixture of Na2U2O7 and UO3, and U4+ as UO2). Chemical impurities were mainly Na (5%), Al (1–2%) and Fe (0.4–4%), and radioactive impurities 90Sr, 144Ce, 137Cs, 155Eu, 125Sb, 106Ru, 239Pu and 60Co at activity levels ranging from 1x109 to 3x104 Bq/g U. Optimized dissolution process parameters (solid/liquid ratio of 1:12, addition of 10 ml 1 M (NH4)2CO3 and 1–2 ml 30% H2O2 per g of residue at 60 °C with a reaction time of one hour), provided a uranium recovery yield of 99%. Characterization of the orangered complex formed during dissolution confirmed it to be mainly [UO2(O2)(CO3)2]4–, which is unstable and slowly dissociates, precipitating less-soluble (NH4)4UO2(CO3)3. The optimized dissolution parameters were applied to real irradiated residue, where three successive leaches were required for complete uranium recovery due to the presence of high iron levels caused by corrosion of the steel canister in which the residue is stored. Uranium purification through dissolution and ion exchange on alumina was determined in these runs, which showed that the chemical impurities iron and aluminium, and the lanthanide fission products as well as 239Pu were efficiently removed. However, dose rates for purified uranium solution were still above tolerable levels, which prompted development of a further purification scheme using inorganic ion exchangers from the zeolite, hydrous oxide, hexacyanoferrate and titanate families. vi Substantial decontamination of uranium was achieved using alumina, Manox A, CsTreat® and SrTreat®, combined with the initial dissolution process and a further purification step involving selective precipitation of contaminants by heating. Only 60Co, 106Ru, 125Sb and 90Sr require further decontamination factors of ca. 50–80 in 5- year-old residue. MCNP modelling for a proposed full-scale batch of 1 kg U yielded a dose rate of 0.1 mSv/h at 50 cm for the purified solution, which is within limits for radiologically controlled areas. To clarify the observed behaviour of uranium and its contaminants during carbonate dissolution and on the ion exchangers studied in this work, speciation modelling was performed using the three computer codes OLI, Visual MINTEQ and JESS, and observations were made about the validity of the modelling predictions by comparison with pertinent literature.
Ph.D. (Chemistry)