Amanda J. Youker

Capture chromatography for Mo-99 recovery from uranyl sulfate solutions: Minimum-column-volume design method

Molybdenum-99 (Mo-99), generated from the fission of Uranium-235 (U-235), is the radioactive parent of the most widely used medical isotope, technetium-99m (Tc-99m). An efficient, robust, low-pressure process is developed for recovering Mo-99 from uranyl sulfate solutions. The minimum column volume and the maximum column length for required yield, pressure limit, and loading time are determined using a new graphical method. The method is based on dimensionless groups and intrinsic adsorption and diffusion parameters, which are estimated using a small number of experiments and simulations. The design is tested with bench-scale experiments with titania columns. The results show a high capture yield and a high stripping yield (95±5%). The design can be adapted to changes in design constraints or the variations in feed concentration, feed volume, or material properties. The graph shows clearly how the column utilization is affected by the required yield, loading time, and pressure limit. The cost effectiveness of various sorbent candidates can be evaluated based on the intrinsic parameters. This method can be used more generally for designing other capture chromatography processes.

A Solution-Based Approach for Mo-99 Production: Considerations for Nitrate versus Sulfate Media

Molybdenum-99 is the parent of Technetium-99m, which is used in nearly 80% of all nuclear medicine procedures. The medical community has been plagued by Mo-99 shortages due to aging reactors, such as the NRU (National Research Universal) reactor in Canada. There are currently no US producers of Mo-99, and NRU is scheduled for shutdown in 2016, which means that another Mo-99 shortage is imminent unless a potential domestic Mo-99 producer fills the void. Argonne National Laboratory is assisting two potential domestic suppliers of Mo-99 by examining the effects of a uranyl nitrate versus a uranyl sulfate target solution configuration on Mo-99 production. Uranyl nitrate solutions are easier to prepare and do not generate detectable amounts of peroxide upon irradiation, but a high radiation field can lead to a large increase in pH, which can lead to the precipitation of fission products and uranyl hydroxides. Uranyl sulfate solutions are more difficult to prepare, and enough peroxide is generated during irradiation to cause precipitation of uranyl peroxide, but this can be prevented by adding a catalyst to the solution. A titania sorbent can be used to recover Mo-99 from a highly concentrated uranyl nitrate or uranyl sulfate solution; however, different approaches must be taken to prevent precipitation during Mo-99 production.

Fission Produced 99Mo without a Nuclear Reactor

99Mo, the parent of the widely used medical isotope 99mTc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of 99Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of 99Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The 99Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq 99Mo production run are presented.

Micro-SHINE Uranyl Sulfate Irradiations at the Linac

Peroxide formation due to water radiolysis in a uranyl sulfate solution is a concern for the SHINE Medical Technologies process in which Mo-99 is generated from the fission of dissolved low enriched uranium. To investigate the effects of power density and fission on peroxide formation and uranyl-peroxide precipitation, uranyl sulfate solutions were irradiated using a 50 MeV electron linac as part of the micro-SHINE experimental setup. Results are given for uranyl sulfate solutions with both high and low enriched uranium irradiated at different linac powers.

Development of Recovery and Purification Processes for Mo-99 from an Accelerator Driven Subcritical Target Solution: Determination of Distribution Coefficients for Competing Components and Micro-SHINE Tracer Column Results

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