Production of Y-86 and other radiometals for research purposes using a solution target system.

INTRODUCTION: Diagnostic radiometals are typically obtained from cyclotrons by irradiating solid targets or from radioisotope generators. These methods have the advantage of high production yields, but require additional solid target handling infrastructure that is not readily available to many cyclotron facilities. Herein, we provide an overview of our results regarding the production of various positron-emitting radiometals using a liquid target system installed on a 13 MeV cyclotron at TRIUMF. Details about the production, purification and quality control of (89)Zr, (68)Ga and for the first time (86)Y are discussed. METHODS: Aqueous solutions containing 1.35-1.65 g/mL of natural-abundance zinc nitrate, yttrium nitrate, and strontium nitrate were irradiated on a 13 MeV cyclotron using a standard liquid target. Different target body and foil materials were investigated for corrosion. Production yields were calculated using theoretical cross-sections from the EMPIRE code and compared with experimental results. The radioisotopes were extracted from irradiated target material using solid phase extraction methods adapted from previously reported methods, and used for radiolabelling experiments. RESULTS: We demonstrated production quantities that are sufficient for chemical and biological studies for three separate radiometals, (89)Zr (Asat = 360 MBq/µA and yield = 3.17 MBq/µA), (86)Y (Asat = 31 MBq/µA and yield = 1.44 MBq/µA), and (68)Ga (Asat = 141 MBq/µA and yield = 64 MBq/µA) from one hour long irradiations on a typical medical cyclotron. (68)Ga yields were sufficient for potential clinical applications. In order to avoid corrosion of the target body and target foil, nitrate solutions were chosen as well as niobium as target-body material. An automatic loading system enabled up to three production runs per day. The separation efficiency ranged from 82 to 99%. Subsequently, (68)Ga and (86)Y were successfully used to radiolabel DOTA-based chelators while deferoxamine was used to coordinate (89)Zr.

Implementation of Multi-Curie Production of (99m)Tc by Conventional Medical Cyclotrons.

UNLABELLED: (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% H2O2. (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.

(44g)Sc production using a water target on a 13MeV cyclotron.

INTRODUCTION: Access to promising radiometals as isotopes for novel molecular imaging agents requires that they are routinely available and inexpensive to obtain. Proximity to a cyclotron center outfitted with solid target hardware, or to an isotope generator for the metal of interest is necessary, both of which can introduce significant hurdles in development of less common isotopes. Herein, we describe the production of 44Sc (t1/2=3.97 h, Eavg,ß⁺=1.47MeV, branching ratio=94.27%) in a solution target and an automated loading system which allows a quick turn-around between different radiometallic isotopes and therefore greatly improves their availability for tracer development. Experimental yields are compared to theoretical calculations. METHODS: Solutions containing a high concentration (1.44-1.55g/mL) of natural-abundance calcium nitrate tetrahydrate (Ca(NO3)2·4 H2O) were irradiated on a 13MeV proton-beam cyclotron using a standard liquid target. (44g)Sc was produced via the 44Ca(p,n)(44g)Sc reaction. RESULTS: (44g)Sc was produced for the first time in a solution target with yields sufficient for early radiochemical studies. Saturation yields of up to 4.6 ± 0.3 MBq/µA were achieved using 7.6 ± 0.3 µA proton beams for 60.0 ± 0.2 minutes (number of runs n=3). Experimental data and calculation results are in fair agreement. Scandium was isolated from the target mixture via solid-phase extraction with 88 ± 6% (n=5) efficiency and successfully used for radiolabelling experiments. The demonstration of the production of 44Sc in a liquid target greatly improves its availability for tracer development.

Radiometals from liquid targets: 94mTc production using a standard water target on a 13 MeV cyclotron.

Solutions containing a high concentration (0.325-0.995 g/ml) of natural-abundance ammonium heptamolybdate tetrahydrate ((NH(4))(6)Mo(7)O(24))·4H(2)O were irradiated at 13 MeV on a proton-beam cyclotron using a standard liquid target. (94m)Tc was produced via the (94)Mo(p,n)(94m)Tc reaction with measured yields of 110±20 MBq for the highest concentrated solution using 5 µA proton beams for 60 min. Saturation yields of up to 40±6 MBq/µA were achieved. Pertechnetate was isolated from the target mixture with 70.9±0.7% efficiency using a solid-phase extraction resin.