At present, most 99 Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world's supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassembly of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration. FIGURE 1.3 (A) Target Insertion Rig and (B) Target Insertion Rig with Target Showing Cooling Channel on Inside of Target. Up to Three Targets Can Be Placed on Each Rig.

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99 Mo, the parent of the widely used medical isotope 99m Tc, 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 99 Mo 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 99 Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The 99 Mo 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 99 Mo production run are presented. Thedaught er of 99 Mo (half-life, 66 h), 99m Tc (half-life, 6 h), is used in more than 45 million diagnostic nuclear medicine procedures annually worldwide, with approximately 16.7 million procedures performed in the United States alone (1). Despite being the largest single user of 99m Tc, the United States currently imports 100% of its supply (2). Current supply chains of 99 Mo rely on aging nuclear reactors, such as the High Flux Reactor in The Netherlands and the National Research Universal reactor in Canada (1). The major U.S. supplier, located in Canada, will cease normal production in late 2016 but will maintain the ability to produce 99 Mo for another 2 y, if a severe shortage occurs (3). The High Flux and National Research Universal reactors have exceeded their initial design lifetimes of 40 y, having been in operation for 55 and 59 y, respectively. In 2009-2010, both reactors were shut down for extended periods of time, causing a severe 99 Mo shortage (4,5). The 99 Mo shortage forced clinicians to ration imaging procedures, which delayed critical diagnostic tests or resulted in older, less effective techniques that in many cases increased the radiation dose to the patient (6). The U.S. 99 Mo market is also fragile because it relies solely on international air transportation, which has been halted in the past due to inclement weather, natural phenomena, flight delays, and terrorist threats (6).The predominant global 99 Mo production route is irradiation of highly enriched uranium (HEU, $20% 235 U) solid targets in nuclear reactors fueled by uranium (2). Other potential 99 Mo production paths include (n,g) 98 Mo and (g,n) 100 Mo; however, both routes require enriched molybdenum material and produce low-specificactivity 99 Mo, which cannot be loaded directly on a commercial 99m Tc generator. The U.S. National Nuclear Security Administration implements the long-standing U.S. policy to minimize and eliminate HEU in civilian applications by working to convert research reactors and medical isotope production facilities to low enriched uranium (LEU, ,20% 235 U) worldwide (7). In 2009, the Global...

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.

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