Rhenium-186 has a half-life (t1/2 = 3.72 days) and emission of both gamma and beta particles that make it very attractive for use as a theranostic agent in targeted radionuclide therapy. 186Re can be readily prepared by the 185Re(n,γ)186Re reac-tion1. However, that reaction results in low specific activity, severely limiting the use of reactor produced 186Re in radiopharmaceuticals. It has previously been shown that high specific activity 186Re can be produced by cyclotron irradiations of 186W with protons and deuterons2,3. In this investigation we evaluated the 186W(d,2n)186Re reaction using thick target irradiations at higher incident deuteron energies and beam currents than previously reported. We elected not to use copper or aluminum foils in the preparation of our 186W targets due to their activation in the deuteron beam, so part of the investigation was an evaluation of an alternate method for preparing thick targets that withstand μA beam currents. Irradiation of 186W. Initial thick targets (~600-1100 mg) were prepared using 96.86% enriched 186W by hydraulic pressing (6.9 MPa) of tungsten metal powder into an aluminum target support. Those thick targets were irradiated for 10 minu-tes at 10 µA with nominal extracted deuteron energies of 15, 17, 20, 22, and 24 MeV. Isolation of 186Re. Irradiated targets were dissolved with H2O2 and basified with (NH4)2CO3 prior to separation using column(s) of ~100–300 mg Analig Tc-02 resin. Columns were washed with (NH4)2CO3 and the rhenium was eluted with ~80˚C H2O. Gamma-ray spectroscopy was per-formed to assess production yields, extraction yields, and radionuclidic byproducts. Recycling target material. When tested on a natural abundance W target, recovery of the oxidized WO4- target material from the resin was found to proceed rapidly with the addition of 4M HCl in the form of hydrated WO3. The excess water in the WO3 was then removed by calcination at 800 °C for 4 hours. This material was found to undergo reduction to metallic W at elevated temperatures (~1550 °C) in a tube furnace under an inert atmosphere (Ar). Quanti-fication of % reduction and composition analyses were accomplished with SEM, EDS, and XRD and were used to characterize and compare both the WO3 and reduced Wmetal products to a sample of commercially available material. Structural enhancement by surface annealing. In some experiments ~1 g WO3 pellets were prepared from Wmetal that had been chemically treated to simulate the target material recovery process described above. Following calcination, the WO3 was allowed to cool to ambient temperature, pulverized with a mortar and pestle and then uniaxially pressed at 13.8 MPa into 13 mm pellets. Conversion of the WO3 back to Wmetal in pellet form was accomplished in a tube furnace under flowing Ar at 1550 °C for 8 hours. Material characterization and product composition analyses were conducted with SEM, EDS, and XRD spectroscopy. Graphite-encased W targets. Irradiations were conducted at 20 μA with a nominal extracted deuteron energy of 17 MeV using thick targets (~750 mg) of natural abundance tungsten metal powder uniaxially pressed into an aluminum target support between layers of graphite pow-der (100 mg on top, 50 mg on the bottom). Targets were then dissolved as previously described and preliminary radiochemical isola-tion yields obtained by counting in a dose calibrator. Although irradiations of W targets were possible at 10 μA currents, difficulties were encountered in maintaining the structural integrity of the full-thickness pressed target pellets under higher beam currents. This led to further investigation of the target design for irradiations conducted at higher beam currents. Comprehensive target material characterization via analysis by SEM, EDS, XRD, and Raman Spectroscopy allowed for a complete redesign of the target maximizing the structural integrity of the pressed target pellet without impacting production or isolation. At the 10 A current, target mass loss following irradiation of an enriched 186W target was < 1 % and typical separation yields in excess of 70 % were observed. Saturated yields and percent of both 183Re (t½ = 70 days) and 184gRe (t½ = 35 days) relative to 186gRe (decay corrected to EOB) are reported in TABLE 1 below. The reason for the anomalously low yield at 24 MeV is unknown, but might be explained by poor beam alignment and/or rhenium volatility during irradiation. Under these irradiation conditions, recovery yields of the W target material from the recycling process were found to be in excess of 90% with no discernable differences noted when compared to commercially available Wmetal and WO3. Conceptually, increasing the structural integrity of pressed WO3 targets by high temperature heat treatment under an inert atmosphere is intriguing. However, the treated pellets lacked both density and structural stability resulting in disintegration upon manipulation , despite the initially encouraging energy dispersive X-ray spectroscopy (EDS) determination that 94.9% percent of the WO3 material in each pellet had been reduced to metallic W. The use of powdered graphite as a target stabi-lizing agent provided successful irradiation of natural abundance W under conditions where non-stabilized targets failed (20 µA at 17 MeV for 10 minutes). Target mass loss following irradiation of a natW target was < 1 % and a separation yield in excess of 97 % was obtained. In conclusion, the theranostic radionuclide 186Re was produced in thick targets via the 186W(d,2n) reaction. It was found that pressed W metal could be used for beam currents of 10 μA or less. For deuteron irradiations at higher beam currents, a method involving pressing W metal between two layers of graphite provides increased target stability. Both target configurations allow high recovery of radioactivity from the W target material, and a solid phase extraction method allows good recovery of 186Re. An effective approach to the recycling of enriched W has been developed using elevated temperature under an inert atmosphere. Further studies are underway with 186W targets sandwiched by graphite to assess 186Re production yields, levels of contaminant radiorhenium, power deposition, and enriched 186W material requirements under escalated irradiation conditions (20 µA and 17 MeV for up to 2 hours).

Introduction The COSTIS system is a commercially available target station for the irradiation of solid targets. Up to 3 targets can be provided for irradiation by a slot system. In standard setup the target can be ejected via a pneumatically driven piston system. The target is then allowed to drop down into an open lead container, which can be closed remotely afterwards. The described procedure is well established and reliable. But the concept is limited to low dose targets and environments. The required entering of the cyclotron vault for manual pick up of the container at the cyclotron and the light 18 mm Pb lead shielding of the container itself cause exposure risk for the personnel after long term irradiations with highly activated cyclotron parts and target. The purpose of this work was the design of an alternative for the pickup and the transport of irradiated targets to minimize the radiation dosage of the personnel during manual handling of the COSTIS-lead container. Principle The new designed transport system still uses the software controlled target ejection function of the COSTIS/IBA-system. With ejection the target capsule is allowed to fall into a PTFE-container. To assure a safe target drop into the PTFE container, the gap between the target guiding plate and the PTFE container is smaller than d/2 of the target capsule. After target ejection the PTFE-container can be transferred remotely from target ejection position (1) to the loading station (2) with a target slide. The loading station allows the transfer of the PTFE container remotely into a lead container (60 mm Pb). Now the vault door is used as carrier of the Pb-container. For this purpose a proper fixture for the Pb-container is mounted at the front side of the vault door and via opening the vault door the container is safely transported out of the vault. Outside the container will be finally closed with a lid and transferred to a trolley for further handling. Due to positioning of the container at a certain altitude together with the deep positioning of the target coin inside of the container, the subsequent closing of the container does not cause significant dosage, a more complicated automatic closing system is not mandatory. After replacement of the lead container further transfers can be executed without entering the vault. For this purpose the exchanged Pb-container is placed at the loading station by closing the vault door and a new PTFE-container will be transferred remotely from a magazine onto the target slide, which again can be re-motely positioned at target ejection position. The magazine of PTFE-Containers holds two replacements in accordance with the maximal capacity of the target slot system of the COSTIS station. The remote system of the transport unit uses redundant feedback signals for a reliable and safe operation. Results and Conclusion The newly implemented transport system allows a significant reduction of the radiation dose during pickup and transport of the irradiated solid targets. No entering of the vault is needed after irradiation. The system is highly reliable due to its redundant and straightforward design (2-fold position switches and photoelectric barriers). Due to fixed attachment points in the vault and at the BTL the mobile unit can be easily removed or mounted. The system is maintenance free and all parts easy accessible. For further handling of the targets lead containers were design to fit in the transfer locks of hot cells. The transfer can be carried out directly from the trolley. Container lid and PTFE container are suited for manipulator handling in hot cells.