Aim/Introduction: Increased use of biomarkers for diagnostics of neuroendocrine tumors and prostate cancer has amplified the clinical demand for 68Ga. For sites with access to a cyclotron, the high price and limited availability/activity output of 68Ge/68Ga generators are strong motivators for production of 68Ga (T1/2=68 min) via the 68Zn(p,n)68Ga-reaction. To expand production capacity over generators and also over liquid cyclotron solution targets the aim of this work is to optimize and automate solid target 68Ga production using enriched 68Zn-foils. Materials and Methods: Enriched (98.80 % 68Zn, 0.46 % 64Zn, 0.43 % 66Zn, 0.29 % 67Zn, 0.02 % 70Zn) zinc foils (CMR, 15.5 mm diameter, 0.10 mm thick) were pneumatically transferred to a solid target system (Comecer EDS/PTS) and irradiated with 25 µA protons (PETtrace, GE Healthcare). Proton energy was degraded to a nominal 12.6 MeV to minimize co-production of long lived 67Ga. Separation of 68Ga from zinc was programmed and automated with a separation module (I.e Comecer, Taddeo PRF). In these initial trials foils were dissolved with 12 M HCl. The solution was diluted to 6 M HCl and passed over a UTEVA resin (Triskem, 100 mg) to trap gallium and to remove zinc and other metals. Gallium was then eluted with 1.2 ml water. Radionuclidic purity (RNP) was determined with an energy and efficiency calibrated HPGe-detector. Results: Irradiation of 136 min (i.e two half-lives) yielded up to 48 GBq of 68Ga (saturation yield = 2.6 GBq/µA) in the foil at end of bombardment (EOB). Transfer (5 min) and gallium isolation (16 min) requires 21 min after EOB. Decay corrected trapping/elution on the UTEVA resin itself exceeds 93 %. Up to 29 GBq of 68Ga was eluted with 1.2 ml of water at end of separation. RNP in the eluate was 99.98 % at EOB. Other gamma lines corresponded to 66Ga (0.010 %) and 67Ga (0.015 %). This equates to an RNP above 98% out to 7.7 hrs post-EOB. Conclusion: This setup produces approximately 20 times more activity than eluates from new generators (1.5 GBq). Solid target production using foils do not require any plating techniques. By using a 0.25 mm thick 68Zn foil, we estimate that production of approximately 140 GBq may be possible. Ongoing ICP-MS and titration studies are underway to assess alignment with current generator formulations. References: none
Aim/Introduction: With increasing clinical demand for gallium-68, the commercial 68Ge/68Ga generators fail to supply sufficient amounts of this short-lived isotope. In this study we develop and evaluate an automated method for multi-Curie production of gallium-68 using solid targetry. Materials and Methods: Gallium-68 was produced by irradiation of an enriched zinc-68 solid target (on silver backing) in an ARTMS QIS on a GE PETtrace cyclotron. Beam currents up to 80 µA were applied for up to 120 min with a proton energy of 13 MeV. After end-of-bombardment (EOB), the targets were automatically transferred to a dissolution cell (ARTMS) in connection to a GE FASTlab 2 synthesizer unit. The targets were dissolved in hot, concentrated HCl followed by radiochemical separation of gallium-68 from the target material on the FASTlab 2. Results: Irradiation was performed using up to 80 µA for 120 min, producing up to 194 GBq (5.24 Ci) of gallium-68 at the end of separation from an expected >370 GBq (>10 Ci) gallium-68 at EOB. The fully automated dissolution/separation was performed in 35 min. Multiple productions were analyzed according to Ph. Eur. Monograph draft1 and found to comply with all tests completed to date. Analysis of metal impurities using ICP-OES is in progress. Importantly, the radionuclidic purity (RNP) was high and allowed for a shelf-life of up to 7 h based on RNP alone. In every instance the radiochemical purity was above 99.9%. Radiolabeling of DOTATATE and PSMA-HBED-11 were performed in high yields (>95%) and in clinically acceptable molar specific radioactivity (≥ 24 MBq/nmol, non-optimized). This indicates a low amount of metallic impurities in the produced gallium-68, i.e., similar to what is observed for the generator-produced isotope. Conclusion: We have developed and evaluated an automated method for production of up to isolated 194 GBq gallium-68 chloride in high radionuclidic and radiochemical purity - expected to be suitable for compounding and subsequent clinical use. References: 1 Gallium (68Ga) Chloride (Accelerator-Produced) Solution for Radiolabelling, Ph. Eur. Monograph draft 3109
Aim/Introduction: Radio-manganese shows promise for a variety of nuclear medical applications including radiolabeling antibodies, monitoring pancreatic beta cell viability , and as dual PET/MR imaging agents . Cyclotron-produced 51Mn or 52gMn must be chemically separated from target materials to render the radiotracers suitable for in vivo use. This process has previously been accomplished only using multiple sequential columns, unwieldy liquid/liquid extractions, or complicated precipitation methods. We studied the TrisKem Actinide (AC) resin to investigate its potential to isolate 51,52gMn from bulk iron, chromium, and common trace metal contaminants. Materials and Methods: Both batch and dynamic column experiments were performed to characterize the AC resin’s affinity for representative masses of various metallic constituents. For batch resin experiments, solutions containing Fe, Mn, Co, Zn, Cu and Cr were loaded onto columns containing 100 mg resin equilibrated with varying concentrations of HNO3 and HCl. These columns were agitated for 30 min. Afterwards, the solution was pushed through the resin, and concentrations of each metal were assayed by Microwave Plasma - Atomic Emission Spectroscopy and HPGe (FWHM @ 1333 keV = 1.6 keV ) to quantify the metal mass adhered to the resin and eluted in solution. Affinity constants were calculated from these data for Fe, Cr, and Mn. Batch experiment results were used to design dynamic column experiments. A column containing 500 mg of resin was equilibrated with 5 mL of 0.05 M HNO3. A dissolved and reconstituted target solution was then loaded onto the column and washed with 46 mL of .05 M HNO3. In this wash, Cr, Fe, and other trace metals eluted from the column while Mn was retained. The Mn was eluted in 2 mL of 5 M HNO3. Results: The results collected from both batch resin experiments and preliminary dynamic column experiments show promising results for a quick and efficient separation chemistry of 51,52gMn from bulk target material. Conclusion: We hope to determine a quantified separation factor from multiple dynamic column experiments, as well as binding efficiency with cheltors such as DOTA, EDTA, and PCTA. These final characterizations of the separation method using TrisKem AC resin will provide insight to the potential application of radio-manganese in nuclear medicine. References:  Fonslet, J. et al. (2017). Optimized procedures for manganese-52: Production, separation and radiolabeling. Applied Radiation and Isotopes.  GJ Topping et al.,“Manganese-52 Positron Emission Tomagraphy Tracer Characterization and Initial Results in Phantoms and In Vivo.”Med Phys (2013).
Aim/Introduction: 161Tb has many attractive properties (half-life of 6.9 d, β- decay energy of 593 keV, conversion and Auger electron emission) for targeted radionuclide therapy with peptides and antibodies . 161Tb can be produced via neutron irradiation of 160Gd in a nuclear reactor, such as BR2 (flux of circa 1014 neutrons cm-2 s-1) at SCK•CEN. The aim of this project was to develop a fast and easy method to purify small amounts (circa 250 MBq) of n.c.a. 161Tb with high specific activity, for radiolabeling studies and in-house chelator design. Materials and Methods: 0.5 mg Enriched 160Gd (98.2%) was loaded in a quartz ampoule under vacuum and irradiated at SCK•CEN. Initial irradiations (ϕ0 = 3.67x1011 neutrons cm-2 s-1) were performed at the BR1 reactor to test the separation procedure using tracer amounts. Following a 2 day cooling time, the ampoules were transported to the lab for processing. High performance ion chromatography (HPIC) was used to separate 161Tb from the target matrix on a strong cation exchange resin by elution with α-hydroxyisobutyric acid (α-HIBA). Gamma spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) were used for 161Tb characterization and quantification. To show that the 161TbCl3 can be used for radiolabeling, we performed some radiolabeling experiments with DOTA derivatives as a proof of concept. Results: With the use of a step-wise gradient elution of α-HIBA 161Tb was separated from the Gd target material. 50 kBq 161Tb (specific activity of 49.2 GBq/mg 161Tb) was collected in one fraction of circa 1 ml. Gamma spectrometry analysis pre- and post-purification showed successful separation of the 161Tb from the 159Gd (363.3 keV γ-emission). 161Tb was characterized and quantified by the analysis of the 25.5 (22%), 48.9 (16%), 57.2 (2%) and 74.6 (10%) keV γ-emissions . ICP-MS analysis of the produced 161Tb is still ongoing. Initial radiolabeling experiments with DOTA derivatives however showed high radiolabeling yields could be achieved with our 161TbCl3 solution. Conclusion: Preliminary results show that a fast and efficient method to purify MBq amounts of n.c.a. 161Tb has been developed. Further irradiations in the BR2 reactor are in progress to validate the procedure, analyse the radiochemical purity and yield by using higher activities and thus leading to regular production of 161Tb for in-house research at SCK•CEN. References:  Kostelnik, T.I. and C. Orvig, Chemical Reviews, 2018. 119, 902-956. Tuli, J.K., Nuclear Data Sheets, 1974. 13, 493-547.2. Tuli, J.K., Nuclear Data Sheets, 1974. 13, 493-547.
Aim/Introduction: 161Tb (T1/2 = 6.89 d) is a therapeutic radiolanthanide which shows similar decay characteristics and chemical behaviour to that of 177Lu. While 177Lu is currently regarded as the “gold standard” of radionuclide therapy, the therapeutic effect of the former may be superior as a result of its co-emission of conversion and Auger electrons . The production of 161Tb was reported previously , however, further development and improvement of the 161Tb purification process took place and will be presented. The product was characterized and the 161Tb purity compared with that of 177Lu. Materials and Methods: Enriched 160Gd oxide targets were irradiated at the SAFARI-1 (South Africa) reactor and the high flux reactor of ILL (France), as well as the spallation-induced neutron source (SINQ) at PSI, Switzerland, using the 160Gd(n,γ)161Gd→161Tb nuclear reaction to produce a no-carrier-added (n.c.a.) product. 161Tb separation from the target material was performed using cation exchange chromatography, while the desired 161Tb was concentrated using extraction chromatography before elution of the final product in a small volume. The pH, radionuclidic and radiochemical purity of 161TbCl3 was determined, with the radiolabeling capacity of 161Tb monitored over a two-week period post processing. The product was assessed metrologically towards future instrument calibration. The 161Tb was characterized and half-life measurements performed, using various forms of detection, to ensure accuracy of activity measurements. Results: Irradiations of enriched 160Gd2O3 targets, followed by chemical separation, resulted in yields of 8-20 GBq 161Tb. The final product was obtained with a >80% separation yield and activity concentration of 11-21 MBq/µL. The radionuclidic purity of 161TbCl3 was ≥99.9% at End of Separation (EOS) with the 160Tb impurity, produced by the 159Tb(n,γ) reaction, determined to be ≤0.007% of the total 161Tb activity at EOS. DOTANOC was labelled with 161Tb at 180 MBq/nmol specific activity at a labelling efficiency of ≥99%. The radiolabelling yield of DOTA with 161Tb was comparable to n.c.a. 177Lu over a two-week period. The half-life measurements of 161Tb are ongoing. Conclusion: High yields of 161TbCl3 in a quantity and quality suitable for high-specific radiolabelling, useful for preclinical and potential clinical application, was produced using a variety of irradiation sources and an innovative chemical separation method. References:  Müller et al. Eur. J. Nucl. Med. Mol. Imaging 2014; 41: 476-85.  Lehenberger et al. Nucl. Med. Biol. 2011; 38: 917-24.
Aim/Introduction: The Auger and conversion electron emitting radionuclide 119Sb (t1/2 = 38.5 h) is a candidate for targeted radionuclide therapy (TRT) due to highly localized energy deposition of its emitted low energy electrons . Antimony-119 can be produced using 119Sn(p,n)119Sb reactions initiated by proton irradiation on small cyclotrons, providing a scalable production route . Our efforts build upon previous work in Sn target development and chemical separation  with a focus on producing radioantimony suitable for chelation-based labeling of biological targeting vectors. Materials and Methods: At the University of Wisconsin Madison, thick, electrodeposited natSn targets  were irradiated with 16 MeV protons using a GE PETtrace, producing various radioisotopes of antimony. Chemical separation techniques explored and optimized included liquid-liquid separation, cation exchange (AG 50W-8X), and extraction chromatography. Chelation reactions with trithiol chelator, designed and synthesized by the Jurrisson Radiochemistry group at the University of Missouri , and Sb-chelate stability in mouse serum were monitored using analytical HPLC. Results: Various Sb radioisotopes were produced from natSn, with 120mSb (t1/2= 5.76 d), 122Sb (t1/2= 2.74 d), and co-produced 117mSn (t1/2= 14.0 d) used as tracers for development and optimization of chemical separation techniques. The cation exchange separation method developed produces a separation factor between Sn and Sb of >104 and resulted in chemical purity and Sb speciation suitable for chelation. Reactions resulting in trithiol chelation of radioantimony isolated via cation exchange chromatography were fast (<30 min) and efficient (>95%). Recycling techniques that enable recovery of enriched 119Sn are being developed and will be discussed. Conclusion: Radioisotopes of antimony were produced via proton irradiation of natSn, chemically separated, and stably chelated, with ongoing studies into the recycling of target material for future application of enriched 119Sn targets for 119Sb production. References:  H. Thisgaard and M. Jensen, “Production of the Auger emitter 119Sb for targeted radionuclide therapy using a small PET-cyclotron,” Appl. Radiat. Isot., 2009. H. Thisgaard, M. Jensen, and D. R. Elema, “Medium to large scale radioisotope production for targeted radiotherapy using a small PET cyclotron,” Appl. Radiat. Isot., 2011. P. Møller and L. P. Nielsen, Advanced Surface Technology, 2nd ed. 2013. A. J. DeGraffenried, Y. Feng, C. L. Barnes, A. R. Ketring, C. S. Cutler, and S. S. Jurrison, “Trithiols and their Arsenic Comounds for Potential Use in Diagnostic and Therapeutic Radiopharmaceuticals,” Nucl Med Biol, vol. 43, no. 5, pp. 288-295, 2016.