Article pubs.acs.org/IECR
A Novel Method for Molybdenum-99/Technetium-99m Recovery via Anodic Carbonate Dissolution of Irradiated Low-Enriched Uranium Metal Foil M. Alex Brown, Artem V. Gelis,* Jeffrey A. Fortner, James L. Jerden, Stan Wiedmeyer, and George F. Vandegrift Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, Illinois 60439, United States S Supporting Information *
ABSTRACT: A new method is presented here for digesting irradiated low-enriched uranium foil targets in alkaline carbonate media to recover 99Mo. This method consists of the electrolytic dissolution of uranium foil in a sodium bicarbonate solution, followed by the precipitation of carbonate, base-insoluble fission products, activation products, and actinides with calcium oxide; most of the molybdenum, technetium, and iodine remain in solution. An electrochemical dissolver and mixing vessel were designed, fabricated, and tested for the processing of a full-sized irradiated foil under ambient pressure and elevated temperature. Over 92% of the fission-induced 99Mo was recovered in a product solution that was compatible with an anion-exchange column for retaining molybdenum and iodine.
1. INTRODUCTION
Caustic digestion of irradiated HEU aluminide dispersion targets is already employed for 99Mo production.8,9 However, the density of a uranium aluminide dispersion target will only allow for doubling of the uranium target. Thus far, densities of 2.5−3 g of U/mL in the fuel meat have been achieved, where current HEU target densities are in the range of 1.3−1.6 g of U/mL.10 Uranium metal foil, with a density of 19.04 g of U/ mL,11 would provide more than a 5-fold factor of uranium and thus increase the amount of fissions per target. The LEU-foil annular target has been developed by Argonne with multiple supporting irradiations and processing demonstrations.8,12 Because uranium metal does not digest at a reasonable rate in sodium hydroxide due to the formation of a passive uranium dioxide layer,13 additional oxidizing agents, such as hydrogen peroxide, are necessary to further oxidize UO2 to the uranyl ion UO22+.15 Excess peroxide, although potentially useful for aged nuclear fuel digestion and reprocessing, could interfere with molybdenum and complicate the separation chemistries. Peroxide is also unstable at the contact with metallic surfaces, and thus its application in this alkali-driven process may not be practical. If the uranium metal can be alternatively oxidized to its hexavalent state (i.e., UO3), it can be easily digested in a carbonate solution. The major fission products such as zirconium, niobium, barium, and light lanthanides either are insoluble in bicarbonate or can be coprecipitated with uranium and neptunium via insoluble calcium metal carbonate complexes. Anionic species of molybdenum and iodine should remain in solution.
The γ-emitting metastable isotope of technetium, Tc, accounts for roughly 30 million annual radiopharmaceutical diagnostic procedures worldwide.1 Owing to its short 6 h halflife, however, commercial production and distribution of 99mTc has been the focus around its longer-lived molybdenum parent isotope 99Mo.2 The current global demand of 99Mo has been met thus far primarily by thermal neutron fission of highly enriched uranium (HEU, enriched in 235U) targets, in which approximately 6% of fissions result in 99Mo. In light of the National Nuclear Security Administration’s (NNSA) Global Threat Reduction Initiative (GTRI), removing, securing, and/ or disposing of high-risk vulnerable nuclear material is a top priority.3 Argonne National Laboratory was tasked by GTRI to develop front-end processes for current producers using HEU targets to convert to LEU (enriched to 1000 Ci 99Mo) would ideally utilize an anode basket fabricated in nickel to avoid corrosion during repeat dissolutions. For cost and demonstration purposes, SS was found to be sufficient for these experiments. A 9 V potential was applied to overcome the resistance of the SS anode and cathode (current × resistance drop) and to accelerate oxidation of uranium(III) to uranium(VI). This overpotential also increased the energy deposition in the solution such that the temperature had to be controlled during digestion to avoid boiling of the electrolyte. Hydrogen gas was generated at the cathode, which may require some special off-gas treatment during large-scale production. Overall, 98% dissolution was completed in less than 4 h, which is reasonable for molybdenum processing and purification. It should be mentioned that this time frame included several periods in which electrolysis was stopped and the foil was visually examined. Increasing the surface area of the uranium
μCi
isotope
dissolver solutionb
Mo-99 U-237 La-140 Ba-140 Zr-95 Nb-95 Nd-147 Np-239 Te-132 I-133 I-131 Ru-103 Rh-105
32.1 16.7 5.6 0.4 1.4 0.2 0.8 10.2 0.7 32.3 1.9 0.3 1.4
Mo separated filter 1.9
product solutionc
% Mo recoveredd
Kd (mL/g)e
25.6 0.7 0.7 ND 0.1 ND 0.1 0.3 0.1 11.1 0.9 ND 0.1
92 ± 3
143.0
834.3 692.1
a
These values have been decay-corrected to the time of irradiation. These values may not represent the true values because of the uncertainty in the solution volume and γ interference. cND = below the detection limit, complications with the peak shape, or
