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Introduction Rhenium-186 (t1/2 = 3.72 d) is very attractive for use as a theranostic agent in targeted radionuclide therapy (Eβ max = 1.072 MeV (> 76.6 %); Eγ = 137.2 keV)1. Previously published investigations of high specific activity 186Re production have utilized the 186W(p,n)186Re or 186W(d,2n)186Re reactions2-5. Our group is interested in the refinement and scale-up of the production of high specific activity 186Re by cyclotron irradiations of 186W with deuterons; including investigations of the most suitable target material. WO3 has been successfully used as a target material in proton irradiations by two other groups4,5. Further, the physical properties of WO3, such as the reported monoclinic with Pc space group, body centered cubic crystal structure6 and melting point of 1473 °C, made for an attractive target material as sintered and other more structurally robust pressed pellet target preparations could be explored. Thus, this study reports on the characterization and suitability of WO3 as a full-thickness target material for the deuteron production of 186Re. Materials and Methods Assessments of WO3 for target material suitability and structural integrity were made on thick targets (~1 g) prepared using both commercially available and converted WO3 by either uniaxially pressing (13.8 MPa) of powdered WO3 into an aluminum target support or by placing sintered WO3 pellets (1105 °C for 12 hours) into an aluminum target support. In some experiments, WO3 pellets were prepared by dissolution of Wmetal with H2O2, then treatment with 1.5 M HCl. The recovered hydrated WO3 was calcinated at 800 °C for 4 hours, allowed to cool to ambient temperature, pulverized with a mortar and pestle, uniaxially pressed at 13.8 MPa into pellets with a 13 mm die, and subsequently sintered in a tube furnace under flowing Ar at 1105 °C for 3, 6, and 12 hours. Material characterization and product composition analyses were conducted with SEM, EDS, XRD, Raman spectroscopy, and visible photoluminescence spectroscopy. Thick WO3 targets were irradiated for 10 min at 10 µA with nominal extracted deuteron energies of 17 MeV. Gamma-ray spectroscopy was per-formed to assess production yields and radionuclidic byproducts at least 24 hours post EOB. Results While the color of the commercially available WO3 is slightly different (dull, pale green) than the brighter more yellow color of the chemically processed WO3, X-ray diffraction spectrometry (XRD) indicated the two samples were virtually identical. Attempts to determine how the duration of the sintering process (at 1105 °C) affects the chemical/physical nature of the pellet yielded surprising results. In contrast to the characteristic annealed appearance of sintered material, grains of the WO3 sample appeared more densely packed, but not sintered to one another as had been seen during higher temperature (1550 °C) reductions of WO3 irrespective of the time interval used. Full-thickness pressed or sintered pellets of WO3 were found to disintegrate upon irradiation with the deuteron beam, allowing for the direct irradiation of the aluminum target body producing 24Na as a contaminant. Upon retrieval of the target support it was observed that the WO3 had vaporized, discoloring the surface of the well in the target support and coating the walls of ~61 cm (24 inches) of the terminal portion of the beamline, which then required decontamination. We believe that these observations are the result of outgassing oxygen species that subsequently reacted with the aluminum target support. While these findings are in sharp contrast with the successful production yields and isolations previously reported by both Shigeta et al. and Fassbender et al., we believe that these differences are attributable to differences in target design (previous studies utilized an en-closed target with cooling in front of and behind the target) necessitated by the configuration of our target station. Conclusions. The physical properties of powdered WO3, including its lower melting point and more suitable compressibility than powdered Wmetal, seemed to enhance the structural integrity of a WO3 pellet (whether pressed or sintered). However, when compared to our recent successes with the use of Wmetal based targets, the disappointing degradation of our WO3 targets when irradiated with the incident deuteron beam has led us to believe that Wmetal is the more viable target material for 186Re production in our facility.

Introduction 100Mo target design is key to commercially viable large scale cyclotron production of 99mTc. The target back plate supporting the 100Mo must be chemically inert to the target dissolution conditions but ideally it should also be able to dissipate the high thermal loads of irradiation, not contaminate target substrate with radionuclidic by-products, and be adequately inexpensive to allow for single use. Aluminum was selected as our target support as it satisfies these requirements. Our process entails rolling 100Mo powder into a foil of desired thickness, and then diffusion bonding [1] the foil onto an aluminum back plate. The 100Mo targets were designed to be 20×80×0.1 mm to match our TR24 cyclotron’s proton beam profile and energy. Efforts are currently underway to scale up the process to allow for simultaneous production of multiple targets at once. Material and Methods The crude enriched 100Mo foil (99.815% enrichment) was made from 100Mo powder using a horizontally mounted rolling mill and an aluminum hopper. The crude foil was rolled repeated-ly, and the space between the rollers gradually reduced until the thickness of the foil was changed from an initial thickness of 0.3 mm to a thickness of 0.1 mm. The rolled 100Mo foil was annealed under reducing atmosphere and then bonded to the aluminum target plate support under inert atmosphere in a heated press at 500 °C. Results and Conclusion By rolling 100Mo foils from powder we were able to produce uniform foils with an average density of > 98 % compared to the maximum theoretical density of 100Mo (n = 5) and thicknesses of roughly 0.1 mm. All foils produced were the desired 20 mm width (i.e. limited by the width of the opening of the hopper) and trimmed to the desired 80 mm length. The annealing process was necessary due to the brittleness of the un-annealed rolled foil and the difference in the thermal expansion coefficients of molybdenum and aluminum which caused un-annealed foils in previous experiments to crack and break off during pressing (n = 10). Surface preparation of the aluminum support plate was also found to play a critical step in the efficiency of the bond, and continuing effort to scale the above de-scribed procedure to mass produce 100Mo tar-gets is ongoing. Targets have undergone preliminary testing to 250 μA.

Difficulties with reproducibility of isolation yields when distilling 211At from irradiated bismuth targets led us to use a “wet chemistry” approach for that process1. The wet chemistry approach has provided 211At isolation yields of ~ 78 % after decay and Bi attenuation corrections2. However, the use of diisopropyl ether (DIPE) in the separation process has made it difficult to reach our goal of automating the 211At isolation. Therefore, we have investigated the use of column materials to simplify the isolation of 211At and remove DIPE from the process. In this investigation we evaluated the use of a strong anion exchange resin (AG1×8), a strong cation exchange resin (AG MP-50) and a polyethylene glycol (PEG)-coated resin for separation of 211At from the bismuth target material. Anion and cation resins AG1×8 and AG MP-50 were obtained from commercial sources. A PEG-coated resin was prepared by reaction of the Merrifield resin with mPEG-OH 2000 in the pres-ence of tBuOK at 80 °C for 3 days, followed by drying under vacuum. Prior to use of the PEG resin, it was soaked in H2O. Resins (400–800 mg) were loaded into polypropylene columns (Applied Separations, Inc.). Column elution studies were conducted with and without reductants (0.75M FeSO4/1M H2SO4 or Na2S2O5) to determine their effect on capture of 211At. After target dissolution in HNO3 (and in most cases subse-quent removal of HNO3 by distillation and redis-solution of solid in 8M HCl), 211At solution was loaded onto the column, then the column was washed with 2M HCl or H2O to separate the Bi, and finally was eluted with strong base to remove the 211At. Initial studies were conducted with stable iodine to determine if reductants were effective in the presence of large amounts of bismuth ions. Studies with AG1×8 used 125I to determine if that radiohalogen could be captured and recovered from the column when eluting with boric acid buffers at pH 5.3, 8.0 or 10, or H2O at pH 7. Capture and recovery of 211At was evaluated under the same conditions. Further studies with AG1×8 involved eluting with 4M H2SO4. A limited study with AG MP-50 resin used 1M HCl as eluant. Studies with PEG-coated columns used 2M HCl, 4M HCl, 8M HCl, 16 M HNO3 and 8M HNO3 as initial (capture) eluants. Strong base (0.2, 1 or 12.5 M NaOH; 15M NH4OH) and 3 or 500 mM tetrabutylammonium bromide (TBAB) were evaluated for removal of 211At from the columns tested. The efficiency for capture of 211At on the AG1×8 column was high (99%) when loading with strong acid, but decreased when using 0.1–0.2M boric acid (69–91 %) buffer. Low 211At capture efficiencies were obtained with AG MP-50 col-umns (15–29%). High 211At capture efficiencies (96–100%) were obtained with PEG-coated resins when loading with 8M HCl or 8M HNO3, irre-spective of whether reductant was in the acid solution. Four column washings (2 mL of 2M HCl each) were required to remove all Bi prior to elution of 211At. No bismuth was detected in solution from the 4th washing in any of the elutions studied. Low (< 6%) recovery of 211At from the AG1×8 columns was obtained using the conditions studied. Good (60–79%) recovery of 211At was obtained from PEG-coated resin using 15M NH4OH. Isolation of the 211At from NH4OH solution was accomplished by distillation. In an initial study 211At distilled before obtaining a dry residue. However, later studies demonstrated that addi-tion of NaOH prior to distillation kept the 211At in the distilling flask. These studies demonstrated that PEG-coated columns could be used to isolate 211At from HNO3-dissolved bismuth targets with good non-optimized (~60%) overall recovery yields. The studies are continuing with optimization of elu-tion conditions and automation of the process.

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).

The anaplastic and invasiveness nature of high-grade gliomas involves a synchronized reorganization of the cytoskeleton, membrane recycling, and focal adhesion to the extracellular matrix. These cellular processes are aided by the obligatory water movement which occurs when inorganic ions are transported across the plasma membrane. In a previous study, the presence of Na +-independent K-Cl cotransporter (KCC) isoforms suggested that these membrane proteins had a role in glioma cell motility [1]. Utilizing a rat glioblastoma cell line which shares many characteristics of high-grade gliomas [2], RT-PCR analysis identified RNA expression of: (1) with-no-lysine (WNK) kinases; (2) oxidative stress response (OSR1) kinase; (3) Ste20-related proline-alanine rich kinase (SPAK); (4) protein phosphatase 1 (PP1); and (5) apoptosis-associated tyrosine kinase (AATYK1). Each of these intracellular signaling molecules has been shown to partially regulate cotransporter activity. Although K + influx in Xenopus laevis oocytes co-injected with KCC1 and AATYK1 cRNA was not different under isosmotic conditions, co-injection of KCC3 and AATYK1 cRNA increased isosmotic K + influx to levels observed under hyposmotic conditions. These results suggest that differential expression of specific regulatory proteins might have a role in cotransporter activity, and in turn, possibly glioblastoma cell motility.

A low-cost and efficient method for producing pre-clinical scale quantities of 44gSc is presented. Production involves proton irradiation of natural unenriched calcium metal followed by rapid separation of radioscandium from the target using hydroxmate functionalized resin.© 2012 American Institute of Physics