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Latest News
Researchers report fastest purification of astatine-211 needed for targeted cancer therapy
Astatine-211 recovery from bismuth metal using a chromatography system. Unlike bismuth, astatine-211 forms chemical bonds with ketones.
In a recent study, Texas A&M University researchers have described a new process to purify astatine-211, a promising radioactive isotope for targeted cancer treatment. Unlike other elaborate purification methods, their technique can extract astatine-211 from bismuth in minutes rather than hours, which can greatly reduce the time between production and delivery to the patient.
“Astatine-211 is currently under evaluation as a cancer therapeutic in clinical trials. But the problem is that the supply chain for this element is very limited because only a few places worldwide can make it,” said Jonathan Burns, research scientist in the Texas A&M Engineering Experiment Station’s Nuclear Engineering and Science Center. “Texas A&M University is one of a handful of places in the world that can make astatine-211, and we have delineated a rapid astatine-211 separation process that increases the usable quantity of this isotope for research and therapeutic purposes.”
The researchers added that this separation method will bring Texas A&M one step closer to being able to provide astatine-211 for distribution through the Department of Energy’s Isotope Program’s National Isotope Development Center as part of the University Isotope Network.
Details on the chemical reaction to purify astatine-211 are in the journal Separation and Purification Technology.
Peter S. Ebey, James M. Dole, Arthur Nobile, Jon R. Schoonover, John Burmann, Bob Cook, Steve Letts, Jorge Sanchez, Abbas Nikroo
Fusion Science and Technology | Volume 49 | Number 4 | May 2006 | Pages 859-864
Technical Paper | Target Fabrication | dx.doi.org/10.13182/FST06-A1214
Articles are hosted by Taylor and Francis Online.
The purpose of the experiments described in this paper was to expose samples of polymeric materials to a mixture of deuterium-tritium (DT) gas at elevated temperature and pressure to investigate the effects (i.e., damage) on the materials. The materials and exposure parameters were chosen to be relevant to proposed uses of similar materials in inertial fusion ignition experiments at the National Ignition Facility. Two types of samples were exposed and tested. The first type consisted of 10 4-lead ribbon cables of fine manganin wire insulated with polyimide. Wires of this type are proposed for use in thermal shimming of hohlraums and the goal of this experiment was to measure the change in electrical resistance of the insulation due to tritium exposure. The second type of sample consisted of 20 planar polymer samples that may be used as ignition capsule materials. The exposure was at 34.5 GPa (5010 psia) and 70°C for 48 h. The change in electrical resistance of the wire insulation will be presented. The results for capsule materials will be presented in a separate paper in this issue.
