Vanadium has been found to be orally active in lowering plasma glucose levels; thus it provides a potential treatment for diabetes mellitus. Bis(maltolato)oxovanadium(IV) (BMOV) is a well-characterized organovanadium compound that has been shown in preliminary studies to have a potentially useful absorption profile. Tissue distributions of BMOV compared with those of vanadyl sulfate (VS) were studied in Wistar rats by using 48V as a tracer. In this study, the compounds were administered in carrier-added forms by either oral gavage or intraperitoneal injection. Data analyzed by a compartmental model, by using simulation, analysis, and modeling (i.e., SAAM II) software, showed a pattern of increased tissue uptake with use of 48V-BMOV compared with 48VS. The highest 48V concentrations at 24 h after gavage were in bone, followed by kidney and liver. Most ingested 48V was eliminated unabsorbed by fecal excretion. On average, 48V concentrations in bone, kidney, and liver 24 h after oral administration of 48V-BMOV were two to three times higher than those of 48VS, which is consistent with the increased glucose-lowering potency of BMOV in acute glucose lowering compared with VS.

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This study demonstrates the efficacy of a three-step (64)Cu pretargeting procedure for PET imaging of apoptosis. Our data also confirm the usefulness of small animal PET to evaluate cancer treatment protocols.

99m Tc is currently produced by an aging fleet of nuclear reactors, which require enriched uranium and generate nuclear waste. We report the development of a comprehensive solution to produce 99m Tc in sufficient quantities to supply a large urban area using a single medical cyclotron. Methods: A new target system was designed for 99m Tc production. Target plates made of tantalum were coated with a layer of 100 Mo by electrophoretic deposition followed by high-temperature sintering. The targets were irradiated with 18-MeV protons for up to 6 h, using a medical cyclotron. The targets were automatically retrieved and dissolved in 30% H 2 O 2 . 99m Tc was purified by solid-phase extraction or biphasic exchange chromatography. Results: Between 1.04 and 1.5 g of 100 Mo were deposited on the tantalum plates. After high-temperature sintering, the 100 Mo formed a hard, adherent layer that bonded well with the backing surface. The targets were irradiated for 1-6.9 h at 20-240 μA of proton beam current, producing up to 348 GBq (9.4 Ci) of 99m Tc. The resulting pertechnetate passed all standard quality control procedures and could be used to reconstitute typical anionic, cationic, and neutral technetium radiopharmaceutical kits. Conclusion: The direct production of 99m Tc via proton bombardment of 100 Mo can be practically achieved in high yields using conventional medical cyclotrons. With some modifications of existing cyclotron infrastructure, this approach can be used to implement a decentralized medical isotope production model. This method eliminates the need for enriched uranium and the radioactive waste associated with the processing of uranium targets.

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Background With increasing clinical demand for gallium-68, commercial germanium-68/gallium-68 ([68Ge]Ge/[68Ga]Ga) generators are incapable of supplying sufficient amounts of the short-lived daughter isotope. In this study, we demonstrate a high-yield, automated method for producing multi-Curie levels of [68Ga]GaCl3 from solid zinc-68 targets and subsequent labelling to produce clinical-grade [68Ga]Ga-PSMA-11 and [68Ga]Ga-DOTATATE. Results Enriched zinc-68 targets were irradiated at up to 80 µA with 13 MeV protons for 120 min; repeatedly producing up to 194 GBq (5.24 Ci) of purified gallium-68 in the form of [68Ga]GaCl3 at the end of purification (EOP) from an expected > 370 GBq (> 10 Ci) at end of bombardment. A fully automated dissolution/separation process was completed in 35 min. Isolated product was analysed according to the Ph. Eur. monograph for accelerator produced [68Ga]GaCl3 and found to comply with all specifications. In every instance, the radiochemical purity exceeded 99.9% and importantly, the radionuclidic purity was sufficient to allow for a shelf-life of up to 7 h based on this metric alone. Fully automated production of up to 72.2 GBq [68Ga]Ga-PSMA-11 was performed, providing a product with high radiochemical purity (> 98.2%) and very high apparent molar activities of up to 722 MBq/nmol. Further, manual radiolabelling of up to 3.2 GBq DOTATATE was performed in high yields (> 95%) and with apparent molar activities (9–25 MBq/nmol) sufficient for clinical use. Conclusions We have developed a high-yielding, automated method for the production of very high amounts of [68Ga]GaCl3, sufficient to supply proximal radiopharmacies. The reported method led to record-high purified gallium-68 activities (194 GBq at end of purification) and subsequent labelling of PSMA-11 and DOTATATE. The process was highly automated from irradiation through to formulation of the product, and as such comprised a high level of radiation protection. The quality control results obtained for both [68Ga]GaCl3 for radiolabelling and [68Ga]Ga-PSMA-11 are promising for clinical use.

Recent clinical results have demonstrated remarkable treatment responses of late-stage cancer patients when treated with alpha-emitting radionuclides such as actinium-225 ( 225 Ac). The resulting intense global effort to produce greater quantities of 225 Ac has triggered a number of emerging technologies to produce this rare, yet important, radionuclide. Accelerator-based methods for increasing global 225 Ac production capacity have focused on the high energy (>100 MeV) proton irradiation of thorium, despite the coproduction of the undesirable 227 Ac byproduct at 0.1−0.3% of the 225Ac activity. We at TRIUMF have developed a process for the production of a 225 Ra/ 225 Ac generator from irradiated thorium that results in an 225 Ac product with reduced 227 Ac content. 225 Ac was separated from irradiated thorium and coproduced radioactive spallation and fission products using a thorium peroxide precipitation method followed by cation exchange and extraction chromatography. Stable and radioactive tracer studies demonstrated the ability of this method to separate Ac from most other elements, providing a directly produced Ac product with measured 227 Ac content of (0.15 ± 0.04)%. A second, indirectly produced Ac product with 227 Ac content of <7.5 × 10 −5 % is obtained by repeating the final extraction chromatography step with the 225 Racontaining fraction. The 225 Ra-derived 225 Ac showed similar or improved quality compared to the initial, directly produced 225 Ac product in terms of chemical purity and radiolabeling capability, the latter of which was comparable with other 225 Ac sources reported in the literature.

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