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Direct fluorination of a tosylate or mesylate precursor has been a wide-spread and reliable way for radio-fluorination. This approach can be difficult to achieve when the precursor cannot be easily obtained or is chemically unstable. A possible alternative method is to radiolabel ethylene ditosylate or 1,3-propanediol di-p-tosylate to form a radiofluorinated synthon. Here we present the automation of a simplified and reliable approach for the two-step fluorination using [ 18 F]FP-TMP, an analog of antibacterial agent trimethoprim. We demonstrate the feasibility of purifying the fluorinated synthon via filtration, and one final HPLC purification on a commercially available Trasis AllinOne module. The overall reaction time for the two-step reaction is around 90 min andthe decay-corrected yield for more than fifty preparations of [ 18 F]FP-TMP is 22 ± 5 % with high radiochemical purity (≥ 90 %) and specific activities (147 ± 107 GBq/μmol). All batches passed pre-established quality control specifications, demonstrating the utility of using this method in tracer syntheses that meet good manufacturing practice (GMP) requirement. This method can be adopted to the syntheses of other radiotracers, such as [ 18 F]FE-TMP, (+)-[ 18 F]F-PHNO and [ 18 F]FFMZ., Competing Interests: Declaration of competing interest Mark Sellmyer and Robert Mach have filed US patent US20180104365A1, assigned to the University of Pennsylvania, on radiotracer derivatives of trimethoprim for medical imaging. Both hold equity in Vellum Biosciences, a corporate entity that is commercializing trimethoprim radiotracers. Hsiaoju Lee has a consulting agreement with Vellum Biosciences. All others have no conflict of interest., (Copyright © 2024 Elsevier Inc. All rights reserved.)

Background: Actinium-225 is one of the most promising radionuclides for targeted alpha therapy. Its limited availability significantly restricts clinical trials and potential applications of 225 Ac-based radiopharmaceuticals., Methods: In this work, we examine the possibility of 225 Ac production from the thermal neutron flux of a nuclear reactor. For this purpose, a target consisting of 1.4 mg of 226 Ra(NO 3 ) 2 (T 1/2  = 1600 years) and 115.5 mg of 90 % enriched, stable 157 Gd 2 O 3 was irradiated for 48 h in the Breazeale Nuclear Reactor with an average neutron flux of 1.7·10 13  cm -2 ·s -1 . Gadolinium-157 has one of the highest thermal neutron capture cross sections of 0.25 Mb, and its neutron capture results in emission of high-energy, prompt γ-photons. Emitted γ-photons interact with 226 Ra to produce 225 Ra according to the 226 Ra(γ, n) 225 Ra reaction. Gadolinium debulking and separation of undesirable, co-produced 227 Ac from 225 Ra was achieved in one step by using 60 g of branched DGA resin. After 225 Ac ingrowth from 225 Ra (T 1/2  = 14.8 d), 225 Ac was extracted from the 226 Ra and 225 Ra fraction using 5 g of bDGA resin and then eluted using 5 mM HNO 3 ., Results: Measured activity of 225 Ac showed that 6(1) kBq or 0.16(3) μCi (1σ) of 225 Ra was produced at the end of bombardment from 0.9 mg of 226 Ra., Conclusion: The developed 225 Ac separation is a waste-free process which can be used to obtain pure 225 Ac in a nuclear reactor., Competing Interests: Declaration of competing interest The authors declare that they have no competing financial interests or personal relationships that could have appeared to influence this work., (Copyright © 2024 Elsevier Inc. All rights reserved.)

Bismuth-213 is a radionuclide of interest for targeted alpha therapy and is supplied via a radiochemical generator system through the decay of 225 Ac. Radionuclide generators employ longer lived "parent" radionuclides to routinely supply shorter-lived "daughter" radionuclides. The traditional 225 Ac/ 213 Bi radiochemical generator relies on an organic cation exchange resin where 225 Ac binds to the resin and 213 Bi is routinely eluted. These resins degrade when they absorb large doses of ionizing radiation (>1 × 10 6  Gy/mg), which has been observed when the loading activity of 225 Ac exceeds 2.59*10 9  Bq (70 mCi). Herein we report the development of an electrochemical generator for the supply of 213 Bi that has the potential to overcome this limitation. Bismuth-213 spontaneously electrodeposits onto nickel foils in 0.1 M hydrochloric acid at 70 °C. Using this method, we were able to plate an average of 73 ± 4 % of the 213 Bi in solution and obtain a final 213 Bi recovery of 65 ± 8 % in 0.1 M citrate pH 4.5 via reverse electrolysis using titanium as the cathode. The recovered 213 Bi had an average radiochemical purity of >99.8 % and was successfully used to radiolabel DOTATATE with an average radiochemical yield of 85.1 % (not optimized)., Competing Interests: Declaration of competing interest The authors declare the following financial interests/personal relationships which may be considered as potential competing interests: Tara Mastren reports financial support was provided by Los Alamos National Laboratory. If there are other authors, they declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 Elsevier Inc. All rights reserved.)

Background: Radioligand therapy using alpha emitters has gained more and more prominence in the last decade. Despite continued efforts to identify new appropriate radionuclides, the combination of 225 Ac/ 213 Bi remains among the most promising. Bismuth-213 has been employed in clinical trials in combination with appropriate vectors to treat patients with various forms of cancer, such as leukaemia, bladder cancer, neuroendocrine tumours, melanomas, gliomas, or lymphomas. However, the half-life of 213 Bi (T ½  = 46 min) implies that its availability for clinical use is limited to hospitals possessing a 225 Ac/ 213 Bi radionuclide generator, which is still predominantly scarce. We investigated a new Ac/Bi generator system based on using the composite sorbent α-ZrP-PAN (zirconium(IV) phosphate as active component and polyacrylonitrile as matrix). The developed 225 Ac/ 213 Bi generator was subjected to long-term testing after its development. The elution profile was determined and the elution yield, the contamination of the eluate with the parent 225 Ac and the contamination of the eluate with the column material were monitored over time., Results: The high activity (75 MBq of parent 225 Ac) generator with a length of 75 mm and a diameter of 4 mm containing the composite sorbent α-ZrP-PAN with a particle size of 0.8 to 1.0 mm as the stationary phase, eluted with a mixture of 10 mM DTPA in 5 mM nitric acid, provided 213 Bi with yields ranging from 77 % to 96 % in 2.8 mL of eluate, with parent 225 Ac contamination in the order of 10 -3  %, up to twenty days of use., Conclusion: All the results of the monitored parameters indicate that the composite sorbent α-ZrP-PAN based separation system for the elution of 213 Bi is a very promising and functional solution., Competing Interests: Declaration of competing interest The authors declare that they have no competing interests., (Copyright © 2024 Elsevier Inc. All rights reserved.)

Objectives: To construct and evaluate a 64Cu production system that minimises the amount of costly 64Ni, radionuclidic impurities and nonradioactive metal contamination and maximises radiochemical and radionuclidic purity and molar activity; and to report analytical and quality control methods that can be used within typical PET radiochemistry production facilities to measure metal ion concentrations and radiometal molar activities., Methods: Low volume was ensured by dissolving the irradiated nickel in a low volume of hydrochloric acid (<1 mL) using the concave gold target backing as a reaction vessel in a custom-built target holder. Removal of contaminating 55Co and nonradioactive trace metals was ensured by adding an intermediate hydrochloric acid concentration step during the conventional ion-exchange elution process. The radionuclidic purity of the product was determined by half-life measurements, gamma spectroscopy and ion radiochromatography. Trace metal contamination and molar activity were determined by ion chromatography., Results and Conclusions: On a small scale, suitable for preclinical research, the process produced typically 3.2 GBq 64Cu in 2 mL solution from 9.4 ± 2.1 mg nickel-64 electroplated onto a gold target backing. The product had high molar activity (121.5 GBq/µmol), was free of trace metal contamination detectable by ion chromatography and has been used for many preclinical and clinical PET imaging applications., (Copyright © 2021 The Author(s). Published by Wolters Kluwer Health, Inc.)

Production of medical radionuclides with ISOL facilities is a unique production method that may provide access to preclinical quantities of some rare and potent radionuclides for nuclear medicine. Particularly attention over the past years was focused on several promising candidates for Targeted Radionuclides Therapy (TRT). With this review, we provide some perspectives of using the TRIUMF ISOL facility (ISAC) to produce medical radionuclides for TRT application and highlight our current effort to collect of 165 Er and 155 Tb for Auger Therapy and SPECT imaging, respectively., (Copyright © 2021 Elsevier Inc. All rights reserved.)

89 Zr-radiolabelled proteins functionalised with desferrioxamine B are a cornerstone of diagnostic positron emission tomography. In the clinical setting, 89 Zr-labelled proteins are produced manually. Here, we explore the potential of using a microfluidic photochemical flow reactor to prepare 89 Zr-radiolabelled proteins. The light-induced functionalisation and 89 Zr-radiolabelling of human serum albumin ([ 89 Zr]ZrDFO-PEG 3 -Et-azepin-HSA) was achieved by flow photochemistry with a decay-corrected radiochemical yield (RCY) of 31.2 ± 1.3% ( n = 3) and radiochemical purity >90%. In comparison, a manual batch photoreactor synthesis produced the same radiotracer in a decay-corrected RCY of 59.6 ± 3.6% ( n = 3) with an equivalent RCP > 90%. The results indicate that photoradiolabelling in flow is a feasible platform for the automated production of protein-based 89 Zr-radiotracers, but further refinement of the apparatus and optimisation of the method are required before the flow process is competitive with manual reactions.

47 Sc is one of the most promising theranostic radionuclides, thanks to its low energy γ-ray emission (159 keV), suitable for single photon emission computed tomography imaging and its intense β - emission, useful for tumour treatment. Despite promising preclinical results, the translation of 47 Sc-therapeutic agents to the clinic is hampered by its limited availability. Among different 47 Sc-production routes currently being investigated, the nat V(p,x) 47 Sc reaction has proved to be of particular interest, thanks to the low-cost and easy availability on the market of nat V material and the diffusion of medium energy proton cyclotrons. However, the cross section of this specific nuclear reaction is quite low and small amounts of Sc-contaminants are co-produced at energies E P ≤ 45 MeV, namely 48 Sc and 46 Sc. The main concern with these Sc-contaminants is their contribution to the patient absorbed dose. For such a reason, the absorbed dose contributions to healthy organs and the effective dose contributions by the three radioisotopes, 48 Sc, 47 Sc and 46 Sc, were evaluated using DOTA-folate conjugate (cm10) as an example of radiopharmaceutical product. Considering as acceptable the limits of 99% for the radionuclidic purity and 10% for the contribution of radioactive Sc-contaminants to the total effective dose after 47 Sc-cm10 injection, it was obtained that proton beam energies below 35 MeV must be used to produce 47 Sc through irradiation of a nat V target.

Carbon-11-labeled phosgene ([ 11 C]phosgene, [ 11 C]COCl 2 ) is a useful labeling agent that connects two heteroatoms by inserting [ 11 C]carbonyl ( 11 C=O) function in carbamates, ureas, and carbonates, which are components of biologically important heterocyclic compounds and functional groups in drugs as a linker of fragments with in vivo stability. Development of 11 C-labeled PET tracers has been performed using [ 11 C]phosgene as a labeling agent. However, [ 11 C]phosgene has not been frequently used for 11 C-labeling because preparation of [ 11 C]phosgene required dedicated synthesis apparatus (not commercially available) and had problems in reproducibility and reliability. In our laboratory, an improved method for synthesizing [ 11 C]phosgene using a carbon tetrachloride detection tube kit in environmental air analysis and the automated synthesis system for preparing [ 11 C]phosgene have been developed in 2009. This apparatus has been used for routine synthesis of 11 C-labeled tracers 1-4 times/week. Using [ 11 C]phosgene we have developed and produced many PET radiotracers containing [ 11 C]urea and [ 11 C]carbamate moieties. In this review, we report the performance of our method for preparing [ 11 C]phosgene, including automated synthesis apparatus developed in house, and the application of [ 11 C]phosgene for development and production of 11 C-labeled PET tracers., (Copyright © 2020 Elsevier Inc. All rights reserved.)

Theranostic strategies involve select radionuclides that allow diagnostic imaging and tailored radionuclide therapy in the same patient. An example of a Food and Drug Administration-approved theranostic pair is the 68 Ga- and 177 Lu-labeled DOTATATE peptides, which are used to image neuroendocrine tumors, predict treatment response, and treat disease. However, when using radionuclides of 2 different elements, differences in the pharmacokinetic and pharmacodynamic profile of the agent can occur. Theranostic agents that incorporate the matched-pair radionuclides of scandium- 43 Sc/ 47 Sc or 44 Sc/ 47 Sc-would guarantee identical chemistries and pharmacologic profiles. The aim of this study was to investigate production of 43,44,47 Sc via proton-induced nuclear reactions on titanium nuclei using a 24-MeV cyclotron. Methods: Aluminum, niobium, and tantalum target holders were used with titanium foils and pressed TiO 2 to produce scandium radionuclides with proton energies of up to 24 MeV. Irradiated targets were digested using NH 4 HF 2 and HCl in a closed perfluoroalkoxy alkane vessel in 90 min. Scandium radionuclides were purified via ion-exchange chromatography using branched N,N,N',N' -tetra-2-ethylhexyldiglycolamide. The titanium target material was recovered via alkali precipitation with ammonia solution. Results: Titanium foil and TiO 2 were digested with an average efficiency of 98% ± 3% and 95% ± 1%, respectively. The typical digestion time was 45 min for titanium foil and 75 min for TiO 2 The average scandium recovery was 94% ± 3%, and the average titanium recoveries from digested titanium foil and TiO 2 after precipitation as TiO 2 were 108% ± 8% and 104% ± 5% of initial mass, respectively. Conclusion: This work demonstrated a robust method for the cyclotron production of scandium radionuclides that could be used with natural or enriched TiO 2 target material., (© 2021 by the Society of Nuclear Medicine and Molecular Imaging.)

Objective: Copper-64 is a useful theranostic radioisotope that is attracting renewed interest from the nuclear medicine community in the recent times. This study aims to demonstrate the utility of research reactors to produce clinical-grade 64 Cu via 63 Cu(n,γ) 64 Cu reaction and use it in the form of [ 64 Cu]CuCl 2 as a radiopharmaceutical for PET imaging of cancer in human patients., Methods: Copper-64 was produced by irradiation of natural CuO target in a medium flux research reactor. The irradiated target was radiochemically processed and detailed quality control analyses were carried out. Sub-acute toxicity studies were carried out with different doses of Cu in Wistar rats. The biological efficacy of the radiopharmaceutical was established in preclinical setting by biodistribution studies in melanoma tumor bearing mice. After getting regulatory approvals, [ 64 Cu]CuCl 2 formulation was clinically used for PET imaging of prostate cancer and glioblastoma patients., Results: Large-scale (~ 30 GBq) production of 64 Cu could be achieved in a typical batch and it was adequate for formulation of clinical doses for multiple patients. The radiopharmaceutical met all the purity requirements for administration in human subjects. Studies carried out in animal model showed that the toxicity due to "cold" Cu in clinical dose of [ 64 Cu]CuCl 2 for PET scans would be negligible. Clinical PET scans showed satisfactory uptake of the radiopharmaceutical in the primary cancer and its metastatic sites., Conclusions: To the best of our knowledge, this is the first study on use of reactor produced [ 64 Cu]CuCl 2 for PET imaging of cancer in human patients. It is envisaged that this route of production of 64 Cu would aid towards affordable availability of this radioisotope for widespread clinical use in countries with limited cyclotron facilities.

Purpose: Thorium-226 (half-life 30.6 m) is a radionuclide of interest for use in targeted alpha therapy applications. Due to its short half-life, 226 Th must be provided through a radionuclide generator system from its parent 230 U (20.8 d). Furthermore, as the half-life of 226 Th is very short, it should be provided in a form that is directly amenable to use in biomedical applications., Methods: A reverse radionuclide generator system was developed employing a DGA extraction chromatography column. A 230 U/ 226 Th parent/daughter solution in equilibrium is added to a DGA column in >6 M HCl. The parent 230 U is eluted first in 0.1 M HNO 3 followed by elution of 226 Th in 0.1 M citrate buffer pH 5., Results: Thorium-226 was recovered from the radionuclide generator column with >96% yield. Greater than 99.5% of the 230 U parent was isolated for reuse in the generator. Long term evaluation over six weeks demonstrated consistent supply of 226 Th with greater than 99.5% radionuclidic purity. The only contaminant found in the final product was 230 U (<0.5%)., Conclusions: The reverse radionuclide generator described herein was shown to be a feasible method for providing 226 Th in high yield, purity and in a chemical form that is amenable for direct use in biomedical applications., (Copyright © 2020 Elsevier Inc. All rights reserved.)

[ 11 C]carbon monoxide ([ 11 C]CO) is a versatile synthon for radiolabeling of drug-like molecules for imaging studies with positron emission tomography (PET). We here report the development of a novel, user-friendly, fully automated, and good manufacturing practice (GMP) compliant low-pressure synthesis module for 11 C-carbonylation reactions using [ 11 C]CO. In this synthesis module, [ 11 C]CO was reliably prepared from cyclotron-produced [ 11 C]carbon dioxide ([ 11 C]CO 2 ) by reduction over heated molybdenum and delivered to the reaction vessel within 7 min after end of bombardment, with an overall radiochemical yield (RCY) of 71%. [ 11 C]AZ13198083, a histamine type-3 receptor ligand, was used as a model compound to assess the functionality of the radiochemistry module. At full batch production conditions (55 μA, 30 min), our newly developed low-pressure 11 C-carbonylation apparatus enabled us to prepare [ 11 C]AZ13198083 in an isolated radioactivity of 8540 ± 1400 MBq (n = 3). The radiochemical purity of each of the final formulated batches exceeded 99%, and all other quality control tests results conformed with specifications typically set for carbon-11 labeled radiopharmaceuticals. In conclusion, this novel radiochemistry system offers a convenient GMP compliant production drugs and radioligands for imaging studies in human subjects., (© 2020 The Authors. Journal of Labelled Compounds and Radiopharmaceuticals published by John Wiley & Sons Ltd.)

Acceptable and reproducible radiochemical purity (RP) for 68 Ga-DOTATATE was difficult to obtain with the NETSPOT kit because the manufacturer instructions lacked details on the heater or needles used. Methods: The drug was prepared in an International Organization for Standardization (ISO) 5 environment. Multiple dry baths and needle types were used to investigate the effects of reaction temperature and metal contamination, respectively. Temperature curves were obtained with a calibrated thermocouple. The influence of the accuracy of the NETSPOT reagent volume and its effect on outcome were investigated. Results: The AccuBlock dry bath required recalibration for the ISO 5 environment; after calibration, the temperature was stable (only ±0.1°C from the set point). When we followed package insert recommendations (dry bath temperature set to 98°C, reaction time of 8 min), the reaction temperature was 90.6°C. When Becton Dickinson needles were used for reconstitution, 15 of 18 runs (83%) did not meet the RP specification. However, B. Braun Medical needles achieved satisfactory and stable RP. When the 68 Ga generator was eluted with 5.0 mL of 0.1 M hydrochloric acid (HCl), only 3.8-3.9 mL of eluate reached the reaction vial; this volume did not impact labeling (final pH was 3.8). The labeling success rate increased markedly if the 68 Ga eluate was passed through a conditioned silica gel cartridge or if no cartridge was used; then, RP was more than 99%. HCl contact with the septum of the labeling vial reduced RP. Conclusion: The needle type and the temperature setting of the dry bath have critical roles in 68 Ga-DOTATATE preparation. The AccuBlock dry bath has excellent stability and accuracy and can be used for reliable preparation. By using a conditioned silica gel cartridge or by eliminating the cartridge altogether, the RP is reliably high and stable., (© 2020 by the Society of Nuclear Medicine and Molecular Imaging.)

INTRODUTION:: [Ga]Ga-prostate specific membrane antigen (PSMA)-11 showed a clear gain in sensitivity for lesion detection in the biological recurrence of prostate cancer as compared to the standard [F]fluorocholine radiopharmaceutical. To meet the strong demand for [Ga]Ga-PSMA-11, we aimed to optimize an automated radiolabeling process by evaluating the influence of different key parameters on radiochemical purity and radiochemical yield., Methods: The radiosynthesis of [Ga]Ga PSMA-11 was performed using a Trasis MiniAio synthesizer and a Ge/Ga GalliaPharm generator supplied by Eckert & Ziegler, Berlin, Germany. Optimized labeling parameters were evaluated by variation of sodium acetate concentrations and temperature of radiolabeling as well as the purification process., Results: For each condition tested, radiochemical purity was higher than 99% in the final vial without batch failure, indicating a robust and fast radiosynthesis process. Radiosynthesis without the solid phase extraction purification process at room temperature in less than 5 min resulted in a radiolabeling efficiency of over 99% and remained stable at least 4 h without manual processing to limit operator radiation exposure., Conclusion: The procedure was completely automated and provided a high radiochemical yield. It can be performed several times a day, facilitating the clinical demand of this radiopharmaceutical.

Introduction: The radioisotopes of bromine are uniquely suitable radiolabels for small molecule theranostic radiopharmaceuticals but are of limited availability due to production challenges. Significantly improved methods were developed for the production and radiochemical isolation of clinical quality 76 Br, 77 Br, and 80m Br. The radiochemical quality of the radiobromine produced using these methods was tested through the synthesis of a novel 77 Br-labeled inhibitor of poly (ADP-ribose) polymerase-1 (PARP-1), a DNA damage response protein., Methods: 76 Br, 77 Br, and 80m Br were produced in high radionuclidic purity via the proton irradiation of novel isotopically-enriched Co 76 Se, Co 77 Se, and Co 80 Se intermetallic targets, respectively. Radiobromine was isolated through thermal chromatographic distillation in a vertical furnace assembly. The 77 Br-labeled PARP inhibitor was synthesized via copper-mediated aryl boronic ester radiobromination., Results: Cyclotron production yields were 103 ± 10 MBq∙μA -1 ∙h -1 for 76 Br, 88 ± 10 MBq∙μA -1 ∙h -1 for 80m Br at 16 MeV and 17 ± 1 MBq∙μA -1 ∙h -1 for 77 Br at 13 MeV. Radiobromide isolation yields were 76 ± 11% in a small volume of aqueous solution. The synthesized 77 Br-labeled PARP-1 inhibitor had a measured apparent molar activity up to 700 GBq/μmol at end of synthesis., Conclusions: A novel selenium alloy target enabled clinical-scale production of 76 Br, 77 Br, and 80m Br with high apparent molar activities, which was used to for the production of a new 77 Br-labeled inhibitor of PARP-1., Advances in Knowledge: New methods for the cyclotron production and isolation of radiobromine improved the production capacity of 77 Br by a factor of three and 76 Br by a factor of six compared with previous methods., Implications for Patient Care: Preclinical translational research of 77 Br-based Auger electron radiotherapeutics, such as those targeting PARP-1, will require the production of GBq-scale 77 Br, which necessitates next-generation, high-yielding, isotopically-enriched cyclotron targets, such as the novel intermetallic Co 77 Se., Competing Interests: Declaration of competing interest No potential conflicts of interest relevant to this article exist., (Copyright © 2019 Elsevier Inc. All rights reserved.)

Introduction: Gallium-68 is an important radionuclide for positron emission tomography (PET) with steadily increasing applications of 68 Ga-based radiopharmaceuticals for clinical use. Current 68 Ga sources are primarily 68 Ge/ 68 Ga-generators, along with successful attempts of 68 Ga production using a cyclotron. This study evaluated cyclotron 68 Ga production and automated separation using expeditiously manufactured solid targets, demonstrates an order of magnitude improvement in yield compared to 68 Ge/ 68 Ga generators, and presents a convenient alternative to existing cyclotron production processes. A comparison of radiolabeling and preclinical PET imaging was performed with both cyclotron and generator produced 68 Ga., Methods: 100 mg enriched 68 Zn (99.3% 68 Zn, 0.48% 67 Zn, 0.1% 66 Zn) pellets pressed on silver discs were bombarded for 20-75 min using 12.5 MeV proton beam energies and 10-30 μA currents. 68 Ga was separated using an automated TRASIS AllinOne synthesizer employing AG 50W-X8 and UTEVA resins. Post-separation recovery of the 68 Zn by electrolysis yielded 76.7 ± 4.3%. Radionuclidic purity of cyclotron-produced 68 Ga was investigated with gamma spectroscopy using a HPGe-detector. Radiolabeling was investigated using the macrocyclic chelator DOTA and the bombesin-derived peptide NOTA-BBN2. PET imaging was performed using [ 68 Ga]Ga-NOTA-BBN2 in a PC3 xenograft model., Results: 600 μA·min fresh and recycled quadruplet 68 Zn target irradiations (n = 8) at 12.5 MeV and 30 μA yielded 13.9 ± 1.0 GBq 68 Ga; 2200 μA·min irradiations (n = 3) yielded 37.5 ± 1.9 GBq 68 Ga. HPGe analysis showed EOB 0.0074% and 0.0084% of total activity of 66 Ga and 67 Ga, respectively. Metal impurities were 0.06 ± 0.03 μg/GBq Zn, 0.13 ± 0.007 μg/GBq Fe, and 0.02 ± 0.01 μg/GBq Al for cyclotron 68 Ga. Cyclotron and 68 Ge/ 68 Ga generator 68 Ga respective DOTA and NOTA-BBN2 labeling incorporations were 99.4 ± 0.0% and 99.3 ± 0.2%, and 90.4 ± 1.5% and 93.0 ± 3.6% determined by radio-thin layer chromatography (radio-TLC). Preclinical PET imaging comparison between generator and cyclotron produced 68 Ga showed identical radiotracer tumor uptake and biodistribution profiles in PC3 tumor bearing mice., Conclusions: Cyclotron 68 Ga production provides highly scalable production with equivalent or superior quality 68 Ga to a 68 Ge/ 68 Ga generator, while providing identical biodistribution and tumor uptake profiles. Our described targetry is simpler and more cost-effective than existing liquid and solid targetry, enabling a turnkey production system for multi-facility distribution of cyclotron produced 68 Ga. The manufacturing simplicity described has potential applications for producing other radiometals such as 44 Sc., Advances in Knowledge and Implications for Patient Care: Our cost-effective method of solid target 68 Ga production can enhance 68 Ga production capabilities to meet the high demand for 68 Ga-radiopharmaceuticals for research and clinical use., (Copyright © 2020 Elsevier Inc. All rights reserved.)

18 O-labeled water (Water- 18 O) is a widely used starting material of 18 F-labeled diagnostic agents in positron emission tomography (PET). Conventionally, Water- 18 O has been separated from other stable oxygen isotope species ( 16 O, 17 O) by water distillation or nitric oxide distillation. However, conventional methods are costly and may have safety issues. In 2004, we developed the first unit of our novel oxygen isotope separation process by cryogenic oxygen distillation to overcome these issues. To meet the needs of the market, we built a second unit in 2013 and a third in 2016. We are now operating three commercially viable separation units with a total capacity of 600 kg of Water- 18 O per year., (© 2019 John Wiley & Sons, Ltd.)

We study the possibility of increasing the specific activity of 99 Mo produced by irradiating molybdenum targets with the use of the Szilard-Chalmers effect. According to the Szilard-Chalmers effect, recoil atoms of the 99 Mo radionuclide can be produced in nuclear reactions and retained in surrounding buffer substance. The objective of our work is to measure the yield of recoil atoms in the buffer as a function of the molybdenum layer thickness. The yield of recoil 99 Mo atoms in the 100 Mo(p,x) 99 Mo nuclear reaction as a function of metallic molybdenum nanolayer thickness has been measured. Measurements were carried out after irradiation of nanolayers in the U-150 cyclotron with 28-MeV protons. Nanolayers of molybdenum with thickness of 30 to 220 nm were produced by magnetron sputtering on sapphire plates. The yield of the mass of 99 Mo recoil atoms was found to be the highest with the molybdenum layer thickness of 80 ± 5 nm. The enrichment factor (EF) has been defined as a ratio of the 99 Mo collector specific activity (SA) to the entire target SA before separation as EF = 1200 ± 150. The specific activity of 99 Mo produced in the collector was estimated to be approximately equal to 25 Ci·g -1 ., (© 2019 John Wiley & Sons, Ltd.)

Objective: This study was aimed to assess technical aspects of fractionation of commonly used cold kits in Nuclear Medicine., Materials and Methods: A total of 90 samples (30 samples each) of technetium-99m methylene diphosphonate ( 99m Tc-MDP), 99m Tc-diethylenetriaminepentaacetic acid (DTPA) and 99m Tc-dimercaptosuccinic acid (DMSA III) were taken on various days. The radiochemical purity was calculated of each fraction of these cold kits by using paper chromatography., Results: The mean value of radiochemical purity of 99m Tc -MDP, 99m Tc -DTPA and 99m Tc-DMSA(III) were calculated as ~ 95.12%, 91.43% and 95.68% and standard deviation (SD) were ~ 5.43, 8.36 and 3.88, respectively. Maximum time in which fractionation procedure completed i.e. time required for preparing the fraction or thawing was 10 minutes. All fractionated aliquots were between 1 and 15 days. Radiopharmaceutical bio-distribution was found to be appropriate during imaging in all samples., Conclusion: Fractionation of cold kits using standardised technique is a time and cost-effective method and does not deteriorate the quality of labelling in commonly used pharmaceuticals in our study. We have used fractionated aliquots up to 3 days of preparation in patients with clinically usable radiochemical purity. Deep frozen fractions can be used up to 15 days in our experience.

Application of microfluidics offers numerous advantages in the field of radiochemistry and could enable dramatic reductions in the cost of producing radiotracers for positron emission tomography (PET). Droplet-based microfluidics, in particular, requires only microgram quantities of expensive precursors and reagents (compared to milligram used in conventional radiochemistry systems), and occupies a more compact footprint (potentially eliminating the need for specialized shielding facilities, i.e. hot cells). However, the reported platforms for droplet radiosynthesis have several drawbacks, including high cost/complexity of microfluidic reactors, requirement for manual intervention (e.g. for adding reagents), or difficulty in precise control of droplet processes. We describe here a platform based on a particularly simple chip, where reactions take place atop a hydrophobic substrate patterned with a circular hydrophilic liquid trap. The overall supporting hardware (heater, rotating carousel of reagent dispensers, etc.) is very simple and the whole system could be packaged into a very compact format (about the size of a coffee cup). We demonstrate the consistent synthesis of [ 18 F]fallypride with high yield, and show that protocols optimized using a high-throughput optimization platform we have developed can be readily translated to this device with no changes or re-optimization. We are currently exploring the use of this platform for routine production of a variety of 18 F-labeled tracers for preclinical imaging and for production of tracers in clinically-relevant amounts by integrating the system with an upstream radionuclide concentrator.

Our recent investigations for the radiosynthesis of [ 18 F]fluoromethyl tosylate have highlighted that choice of quaternary methyl ammonium (QMA) cartridge used during the radiosynthesis can significantly impact the radiochemical yields. Often the details of the QMA cartridge used in fluourine-18 syntheses are not fully described. However, our studies demonstrate that the type, the size, and nature (method by which it has been conditioned) of the QMA cartridge used during the radiosynthesis can make a significant impact in the labelling efficiency. This paper investigates the use of three QMA cartridges and demonstrates that radiochemical yield (decay corrected) of [ 18 F]fluoromethyl tosylate can increase from 46% to 60% by simply changing the QMA cartridge (and leaving all other reagents and labelling conditions exactly the same). These learnings may be applied to improve the radiochemical yields of a number of [ 18 F]-fluorinated tracers (and synthons), where the labelling step is base-sensitive to increase the radiochemical yield, thereby significantly benefiting the radiochemistry and nuclear medicine community. This paper also highlights the necessity of the radiochemistry community to ensure the details of QMA cartridges used in fluorine-18 chemistry are fully and accurately described, since this will improve the translation of radiochemical methods from one laboratory to another., (© 2019 John Wiley & Sons, Ltd.)

The ion-production efficiency of a newly developed singly charged ion source (SCIS) has been investigated to discuss the possibility of it being used in an isotope separation on-line system that provides 11 C ions for heavy-ion cancer therapy with simultaneous verification of the irradiation field using positron emission tomography. The SCIS uses a low-energy hollow electron beam to produce singly charged carbon ions efficiently. To deliver sufficient 11 C ions to the treatment room from a limited amount of 11 C molecules, which are produced from a boron compound target and proton-beam irradiation via the 11 B(p,n) 11 C reaction, the SCIS must have high ion-production efficiency. To realize this high efficiency, the SCIS was designed using a three-dimensional particle-in-cell code in previous work. With the fabricated SCIS, we performed experiments to measure the efficiency of producing CO 2 + ions from nonradioactive 12 CO 2 molecules and C + ions from nonradioactive 12 CH 4 molecules. We found that the SCIS achieved efficiencies of ε C + =4×10 -3 (0.4%) for C + production and ε CO 2 + =0.107 (10.7%) for CO 2 + production.

Introduction: Indium-111 when formulated as indium-111 oxine remains the gold standard for long term cell tracking, whereas radiometals for improved PET applications still have to be established. We here describe the on-cartridge formation of gallium-68, zirconium-89 and copper-64 complexes in small volumes suitable for cell labelling, including labelling of red blood cells (RBC) and white blood cells (WBC) and their biological evaluation in vivo., Methods: Small volumes (1-2 mL) of tracers (oxine, tropolone) were directly prepared on an anion exchange cartridge (Sep-Pak QMA). Cells were radiolabelled and the labelling efficiency and efflux were evaluated. The in vivo biodistribution of copper-64-labelled WBC using [ 64 Cu][Cu(oxinate) 2 ] and [ 64 Cu][Cu(tropolonate) 2 ] was monitored in an infection and inflammation animal model using BALB/c mice., Results: On-cartridge concentration of gallium-68, zirconium-89 and copper-64 enabled formation of oxine and tropolone tracers in small volumes with good yields (≥50%) and quality (extraction ≥90%). Prepared tracers radiolabelled the RBC comparable to indium-111 tracers and in vivo biodistribution of copper-64 labelled WBC showed clear accumulation of cells at the site of infection and inflammation., Conclusions: This on-cartridge preparation method enables simple formation of various PET tracers for cell radiolabelling. Zirconium-89 and copper-64 tracers radiolabelled cells with sufficient stability. Due to their longer half-life this approach could be promising for routine applications where longer evaluation periods for cell tracking are needed., Advances in Knowledge and Implications for Patient Care: This novel approach for on-cartridge concentration and preparation of oxine and tropolone precursors with different positron emitters, in small volume and suitable pH, offers a versatile tool towards cell labelling for preclinical and clinical PET applications., (Copyright © 2019 Elsevier Inc. All rights reserved.)

15 O (half-life, 122 s) is a useful radionuclide for PET applications. Current production of 15 O typically makes use of the 14 N(d,n) 15 O, 15 N(p,n) 15 O, or 16 O(p,pn) 15 O reactions using an accelerator. A novel approach for the production of 15 O is via the 16 O(γ,n) 15 O reaction using an electron linear accelerator. Photonuclear reactions using an electron linear accelerator may allow for feasible and economical production of 15 O compared with the current methods. Methods: In this work, experiments using a repurposed Clinac were conducted using oxygen-containing alumina as a target material to study the production rate of 15 O. Additional studies were conducted using a water target cell. Simulations using Geant4 were conducted to predict the activity and power dissipation in the target. Results: Bremsstrahlung radiation from the electron beam, and consequently 15 O production via photonuclear reactions, is enhanced when a high- Z material, tungsten, is placed in front of the target. The alumina irradiations provided preliminary data to optimize the beam parameters and target configuration. The optimal thickness of tungsten was 1.4 mm for both the simulated and the measured studies of alumina. Simulations of irradiated water targets showed that tungsten thicker than 1.4 mm resulted in fewer photons available to activate the water; thus, a higher current was required to achieve a fixed dose. Alternatively, for a constant tungsten thickness, more power was deposited in the target with increasing beam energy, requiring a lower current to achieve a fixed dose. Actual irradiations of a water target yielded a quantity of 15 O in the water that was consistent with expectations based on irradiations of alumina. Conclusion: Several parameters should be considered regarding the photonuclear production of 15 O for an average patient dose of 1,850 MBq (50 mCi) in 10 mL. This work illustrates a variety of machine parameters capable of achieving a reasonable patient dose. Our simulations show that the power deposited in the target for these parameters is less than that in commercially operated cyclotron targets for the production of 18 F. Thus, this work demonstrates that the photonuclear production of 15 O may be a new production path for this useful radionuclide., (© 2019 by the Society of Nuclear Medicine and Molecular Imaging.)

Past and prospective shortages of medical radioisotopes have driven recent developments in the direct production of 99m Tc via the 100 Mo(p,2n) 99m Tc reaction. The cyclotron-based production method has been shown to successfully produce 99m Tc, however trace impurities present in the enriched molybdenum target can also lead to the unintended creation of other radioisotopes which constitute waste. The isotopic composition of the waste has to be investigated in order to determine how it can be handled, transported and safely stored. In this article, we report which waste radioisotopes are created alongside 99m Tc during target irradiation. Results are based on the gamma spectroscopy of waste produced. Significant complexities in the emission spectra made automated identification of radioisotopes inaccurate; complexities were resolved using a manual radioisotope identification procedure. The impact of target composition, integrated beam current and duration of target irradiation on the waste produced was studied. Results indicate that an average of 0.059  ±  0.003 GBq of waste is generated per 1 GBq of 99m Tc produced. Two-thirds of the total waste activity produced was attributed to 99 Mo (T 1/2   =  66 h) alone, while a total of fifty radioisotopes were found in the waste. Long-lived isotopes (T 1/2   >  2 months) constituted only 1% of the total waste activity at end of beam (EOB). In conclusion, it was determined that the waste generated during cyclotron-based 99m Tc production was acceptably low for routine clinical production.

Background: 44Sc is becoming attractive as a PET radionuclide due to its decay characteristics. It can be produced from 44Ca present in natural calcium with 2.08% abundance., Materials and Methods: The targets were mostly prepared from natural CaCO3 or metallic calcium in the form of pellets. After irradiation they were dissolved in 3 M hydrochloric acid and 44Sc was separated from excess of calcium by precipitation of scandium hydroxide using ammonia. Alternatively, targets were dissolved in 11 M hydrochloric acid and 44Sc was separated by extraction chromatography on UTEVA resin. As the next step, in both processes 44Sc was further purified on a cation exchange resin. Initially, the separation procedures were developed with 46Sc as a tracer. Gamma spectrometry with a high purity germanium detector was used to determine the separation efficiency. Finally, the CaCO3 pellet with 99.2% enrichment in 44Ca was activated with protons via 44Ca(p,n)44Sc nuclear reaction., Results: Altogether twenty two irradiations and separations were performed. The working procedures were developed and the quality of separated 44Sc solution was confirmed by radiolabeling of DOTATATE. The chemical purity of the product was sufficient for preclinical experiments. At the end of around 1 hour proton beam irradiation of CaCO3 pellet with 99.2% enrichment in 44Ca the obtained radioactivity of 44Sc was more than 4.8 GBq., Conclusion: 44Sc can be produced inexpensively with adequate yields and radionuclidic purity via 44Ca(p,n)44Sc nuclear reaction in small cyclotrons. The recovery yield in both investigated separation methods was comparable and amounted above 90%. The obtained 44Sc was pure in terms of radionuclide and chemical purity, as shown by the results of peptide radiolabeling.

[ 11 C] CO is a highly versatile synthon that allows for labeling at carbonyl positions of many molecules by means of transition metal-mediated carbonylation reactions. The intrinsic complexity of carbonylation reactions often requires tedious screening of reaction conditions for obtaining satisfying yields. Herein, a [ 11 C] CO dispending system for performing multiple reactions with a single batch of cyclotron-produced [ 11 C]CO 2 is described. This semiautomated setup allows for more rapid and efficient screening of reactions and reaction conditions compared with the traditional "one beam for one reaction" strategy., (© 2018 John Wiley & Sons, Ltd.)

Purpose: In recent years the use of 68 Ga (t 1/2  = 67.84 min, β + : 88.88%) for the labelling of different PET radiopharmaceuticals has significantly increased. This work aims to evaluate the feasibility of the production of 68 Ga via the 68 Zn(p,n) 68 Ga reaction by proton irradiation of an enriched zinc solution, using a biomedical cyclotron, in order to satisfy its increasing demand., Methods: Irradiations of 1.7 Msolution of 68 Zn(NO 3 ) 2 in 0.2 N HNO 3 were conducted with a GE PETtrace cyclotron using a slightly modified version of the liquid target used for the production of fluorine-18. The proton beam energy was degraded to 12 MeV, in order to minimize the production of 67 Ga through the 68 Zn(p,2n) 67 Ga reaction. The product's activity was measured using a calibrated activity meter and a High Purity Germanium gamma-ray detector., Results: The saturation yield of 68 Ga amounts to (330 ± 20) MBq/µA, corresponding to a produced activity of 68 Ga at the EOB of (4.3 ± 0.3) GBq in a typical production run at 46 µA for 32 min. The radionuclidic purity of the 68 Ga in the final product, after the separation, is within the limits of the European Pharmacopoeia (>99.9%) up to 3 h after the EOB. Radiochemical separation up to a yield not lower than 75% was obtained using an automated purification module. The enriched material recovery efficiency resulted higher than 80-90%., Conclusions: In summary, this approach provides clinically relevant amounts of 68 Ga by cyclotron irradiation of a liquid target, as a competitive alternative to the current production through the 68 Ge/ 68 Ga generators., (Copyright © 2018 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.)

In the process of developing [ 18 F]FBEM coupled target peptide, we have instituted a robust automated synthesis of [ 18 F]FBEM, a sulfhydryl (-SH) site specific agent for radiolabeling of peptides and proteins. The radiosynthesis generated 1.67-3.89 GBq (45.1-105.1 mCi, 7.5-18.8% non-decay corrected yield) of [ 18 F]FBEM from 22.2 GBq (600 mCi) of starting [ 18 F]fluoride with molar activity of 31.8 ± 5.3 GBq/µmol (0.86 ± 0.14 mCi/nmol) (n = 3) at the end of synthesis. Radiochemical purity was greater than 98%, and total synthesis time was ~90 min., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

Aim: This work describes a small-scale production of iodine-124 using a 16.5 MeV cyclotron, and a subsequent validation of the formulated sodium [124I]iodide solution for routinely clinical applications., Methods: Iodine-124 ( 124 I) was produced via the 124 Te(p, n) 124 I reaction using a 16.5 MeV GE PETtrace® cyclotron. Irradiation was performed with a pre-prepared solid target consisting of [ 124 Te]TeO 2 (99.93%) and Al 2 O 3 . Different layer thicknesses, irradiation and extraction parameters were tested. After irradiation at the cyclotron, the shuttle with irradiated material was transferred fully automatically via a tube system to the Comecer ALCEO ® Halogen 2.0 extraction unit. Iodine-124 was subsequently extracted in form of sodium [ 124 I]iodide ([ 124 I]NaI) in 0.05 N aqueous NaOH solution, followed by reconstitution and validation for preclinical and clinical uses., Results: Good result was achieved using a beam degradation foil of 500 µm thickness in combination with beam currents between 10 and 15 µA. Under these conditions, up to 150 MBq no-carrier-added [ 124 I]NaI was obtained after a 2 h irradiation time in less than 500 µl 0.05 N NaOH. Isolation of [ 124 I]NaI, including evaporation and extraction at the ALCEO ® Halogen EVP unit was accomplished in 90 min 24 h after production (irradiation), the amount of iodine-123 as assessed by gamma-ray spectroscopy was less than 1.5%. The undesirable iodine-125 was not detectable by gamma spectroscopy. The extracted [ 124 I]NaI could be used directly for radiolabeling purposes, and after buffering with phosphate buffered saline (PBS) and sterile filtration for clinical applications., Conclusions: Through the optimized conditions for irradiation and extraction, iodine-124 was produced in good radiochemical yields and high radionuclide purity. The generated injectable [ 124 I]NaI solution was sterile, non-pyrogenic and ready for preclinical and clinical applications after a sterile filtration through a 0.22 µm membrane filter., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

A novel, rapid, and automated loop method for the synthesis of [ 11 C]nicotine was developed and optimized. The method involves, a reaction of the precursor, (+) nornicotine or (-) nornicotine, with a gas-phase produced [ 11 C]CH 3 I in an 800 µL loop at 75 °C for 5 min followed by a semi-preparatory Reversed-Phase High-Performance Liquid Chromatography (RP-HPLC) purification. The optimized synthesis and purification process was complete in < 30 min and produced [ 11 C]nicotine with > 99.9% Radiochemical Purity (RCP), no [ 11 C]CH 3 I, no (+) nornicotine, 105 mCi/µmole specific activity, 7.0 - 7.2 pH, and 16.6% ethanol. The current method can be optimized, to reduce the ethanol content (<10%), and can be translated to a cGMP production of [ 11 C]nicotine for human clinical trials., (Copyright © 2018 Elsevier Ltd. All rights reserved.)

Gas target systems have been used for decades on cyclotrons to produce radionuclides for medical imaging. However, the activity recovered from such targets is often lower than its theoretically predicted value. Past research has suggested that nuclide interactions with the walls of the target body may play a key role in the loss of recoverable radionuclide activity. Here, we consider gas targets and modify the standard radionuclide production equation by adding a loss term representing radionuclides depositing on the walls of the target. We derive the form of the deposition term based on a simple adsorption model which is then linearized by solving for leading order terms. The resulting production equation uses one fitting parameter to give an estimate of the recoverable activity produced in a target system, taking adsorption into account. The model is then fit to six data series, taken in-house and reported in the literature and a parity plot compares model predictions to experimental data. The model is able to better track the data than any previous models, and points towards a phenomenological understanding of adsorption in target systems.

The ISOLPHARM (ISOL technique for radioPHARMaceuticals) project is dedicated to the development of high purity radiopharmaceuticals exploiting the radionuclides producible with the future Selective Production of Exotic Species (SPES) Isotope Separation On-Line (ISOL) facility at the Legnaro National Laboratories of the Italian National Institute for Nuclear Physics (INFN-LNL). At SPES, a proton beam (up to 70 MeV) extracted from a cyclotron will directly impinge a primary target, where the produced isotopes are released thanks to the high working temperatures (2000 °C), ionized, extracted and accelerated, and finally, after mass separation, only the desired nuclei are collected on a secondary target, free from isotopic contaminants that decrease their specific activity. A case study for such project is the evaluation of the feasibility of the ISOL production of 64 Cu and 67 Cu using a zirconium germanide target, currently under development. The producible activities of 64 Cu and 67 Cu were calculated by means of the Monte Carlo code FLUKA, whereas dedicated off-line tests with stable beams were performed at LNL to evaluate the capability to ionize and recover isotopically pure copper.

We report the single-step synthesis of radioactive gold nanoparticles with an activity and size appropriate for potential use in cancer treatment and diagnosis. Methods: A solution of 2 mM gold chloride (HAuCl 4 ⋅3H 2 O), 1 mM polyvinylpyrrolidone (molecular weight, 360,000), and 60 mM 2-propanol was prepared in deionized water. Seven different samples of the solution were irradiated in a neutron flux of 7.45 × 10 12 n/cm 2 ⋅s in a research reactor for 0.5, 1, 3, 5, 10, 30, or 60 min. The resulting nanoparticles were characterized for morphology and chemical composition using a transmission electron microscope and ImageJ. Results: The obtained nanoparticles were 3-450 nm in size. The average size depended on the length of irradiation, with a longer irradiation producing smaller nanoparticles. Irradiation for 10 min produced nanoparticles with characteristics suitable for potential cancer treatment and diagnosis (average size, 50 nm; activity, 6.85 MBq/mL). Conclusion: Direct production of chemically stable radioactive gold nanoparticles was successfully accomplished using the Missouri University of Science and Technology reactor. The nanoparticles had physical and radioactive characteristics potentially useful for cancer treatment and diagnosis., (© 2018 by the Society of Nuclear Medicine and Molecular Imaging.)

Introduction: Production of fluorine-18-labeled radiopharmaceuticals is always associated with the varying levels of the same compound containing stable fluorine-19. In practice, this affects the molar activity (A m ), defined as amount of radioactivity divided by the molar quantity (Bq/mol). We have focused on studying how the material of the transport tubing connecting the cyclotron target chamber to the synthesis device affects the concentration of fluoride in the water arriving to the reaction vessel and subsequently the A m of the fluorine-18 labeled radiopharmaceuticals produced., Methods: Batches of irradiated and non-irradiated water were analyzed for fluoride content after being transported via non-fluorinated (PEEK, PP) and fluorinated (PTFE, ETFE) tubing or using no tubing at all. A m for the [ 18 F]fluoride was determined and compared with the A m of [ 18 F]fluciclatide, synthesized from the same [ 18 F]fluoride containing batches of water., Results: Significantly higher concentrations of fluoride were seen in irradiated water that was transported in fluorinated tubing compared to non-irradiated water transported in tubing of the same material. This elevation of fluoride concentration is presumably caused by the interaction of ionizing radiation with the fluorinated tubing used between the target chamber and hot cell. Likewise, a significant difference was seen for PEEK tubing (non-fluorinated). This could be due to the fact that fluorine containing compounds are used in the manufacture of PEEK. When using fluorinated tubing for transport of the irradiated water, the resulting fluciclatide concentrations were significantly higher compared to when using non-fluorinated tubing. No significant difference was seen between fluciclatide concentrations when PTFE or ETFE tubing was compared to each other. Using no tubing resulted in lowest fluciclatide concentration., Conclusions: Fluorinated tubing is a source of stable fluoride, and A m can be increased by using non-fluorinated transport tubing. Of all the tubing materials studied PP is preferred., (Copyright © 2018 Elsevier Inc. All rights reserved.)

Introduction: The remarkable stability of the 89 Zr-DOTA complex has been shown in recent literature. The formation of this complex appears to require 89 Zr-chloride as the complexation precursor rather than the more conventional 89 Zr-oxalate. In this work we present a method for the direct isolation of 89 Zr-chloride from irradiated nat Y foils., Methods: 89 Zr, 88 Zr, and 88 Y were prepared by 16 MeV proton irradiation of nat Y foils and used for batch-extraction based equilibrium coefficient measurements for TBP and UTEVA resin. Radionuclidically pure 89 Zr was prepared by 14 MeV proton-irradiation of nat Y foils. These foils were dissolved in concentrated HCl, trapped on columns of TBP or UTEVA resin, and 89 Zr-chloride was eluted in <1 mL of 0.1 M HCl. For purposes of comparison, conventionally-isolated 89 Zr-oxalate was converted to 89 Zr-chloride by trapping, rinsing, and elution from a QMA cartridge into 1 M HCl. Trace metal analysis was performed on the resulting 89 Zr products., Results: Equilibrium coefficients for Y and Zr were similar between UTEVA and TBP resins across all HCl concentrations. K d values of <10 -1  mL/g were observed for Y across all HCl concentrations. K d values of >10 3  mL/g were observed at HCl concentrations >9 M for Zr, falling to K d values of <10 0  mL/g at low HCl concentrations. 89 Zr-chloride was recovered from small columns of TBP in <1 mL of 0.1 M HCl with an overall recovery efficiency of 89 ± 3% (n = 3). An average Y/Zr separation factor of 1.5 × 10 5 (n = 3) was obtained. Trace metal impurities, notably Fe, were higher in TBP-isolated 89 Zr-chloride compared with 89 Zr-chloride prepared using the conventional two-step procedure., Conclusion: TBP-functionalized resin appears promising for the direct isolation of 89 Zr-chloride from irradiated nat Y targets. Excellent 89 Zr recovery efficiencies were obtained, and chemical purity was sufficient for proof-of-concept chelation studies., (Copyright © 2018 Elsevier Inc. All rights reserved.)

High specific activity is often a significant requirement for radiopharmaceuticals. To achieve that with fluorine-18 ( 18 F)-labeled probes, it is mandatory to start from no-carrier-added fluoride and to reduce to a minimum the amount of precursor in order to decrease the presence of any pseudocarrier. In the present study, a feasible and efficient method for microscale one-pot radiosynthesis of 18 F-labeled probes is described. It allows a substantial reduction in precursor, solvent, and reagents, thus reducing also possible side reaction in the case of base-sensitive precursors. The method is based on the use of a small amount of Kryptofix 2.2.2/potassium [ 18 F]fluoride in MeOH (K.222/K[ 18 F]F-MeOH) obtained using Oasis MAX and MCX cartridges. Five methods, differing in terms of MeOH evaporation and precursor addition, for the radiosynthesis of [ 18 F]fallypride and [ 18 F]FET in ≤50-μL scale, were examined and evaluated. The method using the addition of DMSO to the K.222/K[ 18 F]F-MeOH solution prior to MeOH evaporation is proposed as a versatile procedure for feasible one-pot 10- to 20-μL scale radiosyntheses. This method was successfully applied also to the radiosynthesis of [ 18 F]FES, [ 18 F]FLT, and [ 18 F]FMISO, with radiochemical yields comparable with those reported in the literature. Purification of a crude product by an analytical HPLC column was also demonstrated., (Copyright © 2018 John Wiley & Sons, Ltd.)

Background: Peptides labeled with positron-emitting isotopes are emerging as a versatile class of compounds for the development of highly specific, targeted imaging agents for diagnostic imaging via positron-emission tomography (PET) and for precision medicine via theranostic applications. Despite the success of peptides labeled with gallium-68 (for imaging) or lutetium-177 (for therapy) in the clinical management of patients with neuroendocrine tumors or prostate cancer, there are significant advantages of using fluorine-18 for imaging. Recent developments have greatly simplified such labeling: in particular, labeling of organotrifluoroborates via isotopic exchange can readily be performed in a single-step under aqueous conditions and without the need for HPLC purification. Though an automated synthesis has not yet been explored, microfluidic approaches have emerged for 18 F-labeling with high speed, minimal reagents, and high molar activity compared to conventional approaches. As a proof-of-concept, we performed microfluidic labeling of an octreotate analog ([ 18 F]AMBF 3 -TATE), a promising 18 F-labeled analog that could compete with [ 68 Ga]Ga-DOTATATE with the advantage of providing a greater number of patient doses per batch produced., Methods: Both [ 18 F]AMBF 3 -TATE and [ 68 Ga]Ga-DOTATATE were labeled, the former by microscale methods adapted from manual labeling, and were imaged in mice bearing human SSTR2-overexpressing, rat SSTR2 wildtype, and SSTR2-negative xenografts. Furthermore, a dosimetry analysis was performed for [ 18 F]AMBF 3 -TATE., Results: The micro-synthesis exhibited highly-repeatable performance with radiochemical conversion of 50 ± 6% (n = 15), overall decay-corrected radiochemical yield of 16 ± 1% (n = 5) in ~40 min, radiochemical purity >99%, and high molar activity. Preclinical imaging with [ 18 F]AMBF 3 -TATE in SSTR2 tumor models correlated well with [ 68 Ga]Ga-DOTATATE. The favorable biodistribution, with the highest tracer accumulation in the bladder followed distantly by gastrointestinal tissues, resulted in 1.26 × 10 -2  mSv/MBq maximal estimated effective dose in human, a value lower than that reported for current clinical 18 F- and 68 Ga-labeled compounds., Conclusions: The combination of novel chemical approaches to 18 F-labeling and microdroplet radiochemistry have the potential to serve as a platform for greatly simplified development and production of 18 F-labeled peptide tracers. Favorable preclinical imaging and dosimetry of [ 18 F]AMBF 3 -TATE, combined with a convenient synthesis, validate this assertion and suggest strong potential for clinical translation., (Copyright © 2018 Elsevier Inc. All rights reserved.)

This paper presents the irradiation and processing of high-current 100 Mo targets at the University of Alberta (UofA) in a GMP compliant setting. For purpose of comparison with a second production facility, additional studies at Centre Hospitalier Universitaire de Sherbrooke (CHUS) are also described., Introduction: More than 70% of today's diagnostic radiopharmaceuticals are based on 99m Tc, however the conventional supply chain for obtaining 99m Tc is fragile. The aim of this work was to demonstrate reliable high yield production and processing of 99m Tc with medium-energy, high-current, cyclotrons., Methods: We used two cyclotrons (TR-24, Advanced Cyclotron Systems, Inc) for irradiations with 22 MeV or 24 MeV incident energy and 400 μA current up to a maximum of 6 h. The irradiated 100 Mo was dissolved using peroxide, basified using ammonium carbonate, and purified using a PEG-based solid phase extraction technique., Results: High-yield productions with 22 MeV (400 μA, 6 h) yielded an average isolated [ 99m Tc]TcO 4 - yield of 878 GBq ± 99 GBq (23.7 Ci ± 2.7 Ci) decay corrected to EOB, n = 8 (isolated saturation yield: 4.36 ± 0.49 GBq/μA). Irradiations with 24 MeV (400 μA, 6 h) resulted in an average isolated [ 99m Tc]TcO 4 - yield of 993 GBq ± 100 GBq (26.8 Ci ± 2.7 Ci) decay corrected to EOB, n = 7 (isolated saturation yield: 4.97 ± 0.50 GBq/μA). These yields corresponds to 600-700 GBq (16-19 Ci) of [ 99m Tc]TcO 4 - at release (i.e. 3 hour post-EOB). For all tested batches, the QC results were within the recently published specifications in the European Pharmacopoeia., Conclusion: Reliable near-TBq production yields for 99m Tc can be obtained using medium-energy cyclotrons., Advances in Knowledge and Implications for Patient Care: This work presents evidence that medium-energy high-current cyclotrons can provide high yields of [ 99m Tc]TcO 4 - with radionuclidic impurities levels within the specifications of the existing European Pharmacopoeia monograph, indicating that this technology can have a share in the future 99m Tc supply market., (Copyright © 2018 Elsevier Inc. All rights reserved.)

Introduction: Copper-67 is an attractive beta emitter for targeted radionuclide therapy. However, the availability of 67 Cu limits its potential use in a wide range of applications. In this study, we propose an easy small-scale production of 67 Cu using 64 Ni target for a preclinical study., Methods: 67 Cu was produced from an electrodeposited 64 Ni target via the 64 Ni(α, p) 67 Cu-reaction with a 36 MeV alpha beam at 15 eμA (electrical microampere) conducted for 7 h. The chemical separation process of 67 Cu from the 64 Ni target was performed following by our routine procedure of 64 Cu production using cation exchange resin, AG50W-X8, with minor modification. The target and its holder were redesigned in the preparation., Results: The 67 Cu product was obtained with a yield of 55 ± 10 MBq at the end of bombardment (EOB), and the yield was 527 ± 96 kBq/μAh at the EOB. The copper impurity in the product was low (0.71 ± 0.21 μg) and the product was suitable for a preclinical study., Conclusions: We produced 67 Cu with sufficient activity and quality for a preclinical study using a 64 Ni-target. This production method also showed advantages as a routine method, i.e., shorten the processing time, reducing the radiation exposure and ready target recycling, when compared with that of a conventional Zn-target used for 67 Cu production., (Copyright © 2017. Published by Elsevier Inc.)

Intensive efforts were undertaken during the last few decades for the separation of cyclotron-produced 99m Tc from 99 Mo and new papers have been published on this topic since the last review [1]. In the future the cyclotron-based methods can replace reactor-based technology in producing this medical radioisotope and the nuclear reaction 100 Mo(p,2n) 99m Tc appears to be the most worthwhile approach. New ways of producing of 99m Tc require efficient separation methods. Several strategies for separation of 99m Tc from 99 Mo have been already developed. The advantages, disadvantages and technical challenges toward application potential of investigated methods to separate 99m Tc from irradiated 100 Mo target are discussed. These methods include column chromatography, solvent extraction, chemical precipitation and thermochromatography., (Copyright © 2017 Elsevier Inc. All rights reserved.)

Following our previous work on the production of radiometals, such as 64 Cu and 68 Ga, through the irradiation of liquid targets using a medical cyclotron, we describe in this paper a technique to produce 61 Cu through the irradiation of natural zinc using a liquid target. The proposed method is very cost-effective, as it avoids the use of expensive enriched material, and is fast, as a purified solution of 61 CuCl 2 is obtained in less than 30 min after the end of beam. Considering its moderate half-life of 3.33 h and favourable decay properties as a positron emitter, 61 Cu is a very attractive nuclide for the labelling of PET tracers for pre-clinical and clinical use with PET as well as to support the intense R&D programmes being carried out worldwide by taking advantage of the rich and versatile chemistry of copper.

We report herein a modification to a component of the GE radiochemistry system that increases the yield and reliability of [ 11 C]CN - production from [ 11 C]carbon dioxide., (Copyright © 2017 John Wiley & Sons, Ltd.)

Introduction: The production of 18 F-radiotracers using continuous flow microfluidics is under-utilized due to perceived equipment limitations. We describe the dose-on-demand principle, whereby the back-to-back production of multiple, diverse 18 F-radiotracers can be prepared on the same day, on the same microfluidic system using the same batch of [ 18 F]fluoride, the same microreactor, the same HPLC column and SPE cartridge to obtain a useful production yield., Methods: [ 18 F]MEL050, [ 18 F]Fallypride and [ 18 F]PBR111 were radiolabeled with [ 18 F]fluoride using the Advion NanoTek Microfluidic Synthesis System. The outlet of the microreactor was connected to an automated HPLC injector and following the collection of the product, SPE reformulation produced the 18 F-radiotracer in <10% ethanolic saline. A thorough automated cleaning procedure was implemented to ensure no cross-contamination between radiotracer synthesis., Results: The complete productions for [ 18 F]MEL050 and [ 18 F]Fallypride were performed at total flow rates of 20μL/min, resulting in 40±13% and 25±13% RCY respectively. [ 18 F]PBR111 was performed at 200μL/min to obtain 27±8% RCY. Molar activities for each 18 F-radiotracer were >100GBq/μmol and radiochemical purities were >97%, implying that the cleaning procedure was effective., Conclusions: Using the same initial solution of [ 18 F]fluoride, microreactor, HPLC column and SPE cartridge, three diverse 18 F-radiotracers could be produced in yields sufficient for preclinical studies in a back-to-back fashion using a microfluidic system with no detectable cross-contamination., (Crown Copyright © 2017. Published by Elsevier Inc. All rights reserved.)

Scandium-44g (half-life 3.97h) shows promise for application in positron emission tomography (PET), due to favorable decay parameters. One of the sources of 44g Sc is the 44 Ti/ 44g Sc generator, which can conveniently provide this radioisotope on a daily basis at a diagnostic facility. Titanium-44 (half-life 60.0 a), in turn, can be obtained via proton irradiation of scandium metal targets. A substantial 44 Ti product batch, however, requires high beam currents, long irradiation times and an elaborate chemical procedure for 44 Ti isolation and purification. This study describes the production of a combined 175MBq (4.7mCi) batch yield of 44 Ti in week long proton irradiations at the Los Alamos Isotope Production Facility (LANL-IPF) and the Brookhaven Linac Isotope Producer (BNL-BLIP). A two-step ion exchange chromatography based chemical separation method is introduced: first, a coarse separation of 44 Ti via anion exchange sorption in concentrated HCl results in a 44 Tc/Sc separation factor of 10 2 -10 3 . A second, cation exchange based step in HCl media is then applied for 44 Ti fine purification from residual Sc mass. In summary, this method yields a 90-97% 44 Ti recovery with an overall Ti/Sc separation factor of ≥10 6 ., (Copyright © 2017 Elsevier Inc. All rights reserved.)

An increasing interest in zirconium-89 ( 89 Zr) can be attributed to the isotope's half-life which is compatible with antibody imaging using positron emission tomography (PET). The goal of this work was to develop an efficient means of production for 89 Zr that provides this isotope with high radionuclidic purity and specific activity. We investigated the irradiation of yttrium sputtered niobium coins and compared the yields and separation efficiency to solid yttrium coins. The sputtered coins were irradiated with an incident beam energy of 17.5MeV or 17.8MeV providing a degraded transmitted energy through an aluminum degrader of 12.5MeV or 12.8MeV, respectively, with various currents to determine optimal cyclotron conditions for 89 Zr production. Dissolution of the solid yttrium coin took 2h with 50mL of 2M HCl and dissolution of the sputtered coin took 15-30min with 4mL of 2M HCl. During the separation of 89 Zr from the solid yttrium coins, 77.9 ± 11.2% of the activity was eluted off in an average of 7.3mL of 1M oxalic acid whereas for the sputtered coins, 91 ± 6% was eluted off in an average of 1.2mL of 1M oxalic acid with 100% radionuclidic purity. The effective specific activity determined via DFO-SCN titration from the sputtered coins was 108±7mCi/μmol as compared to 20.3mCi/μmol for the solid yttrium coin production. ICP-MS analysis of the yttrium coin and the sputtered coins showed 99.99% yttrium removed with 178μg of yttrium in the final solution and 99.93-100% of yttrium removed with remaining range of 0-42μg of yttrium in the final solution, respectively. The specific activity calculated for the solid coin and 3 different sputtered coins using the concentration of Zr found via ICP-MS was 140±2mCi/μmol, 300±30mCi/μmol, 410±60mCi/μmol and 1719±5mCi/μmol, respectively. Labeling yields of the 89 Zr produced via sputtered targets for 89 Zr- DFO-trastuzumab were >98%. Overall, these results show the irradiation of yttrium sputtered niobium coins is a highly effective means for the production of 89 Zr., (Copyright © 2017 Elsevier Inc. All rights reserved.)

Introduction: The availability of 211 At for targeted alpha therapy research can be increased by the 211 Rn/ 211 At generator system, whereby 211 At is produced by 211 Rn electron capture decay. This study demonstrated the feasibility of using generator-produced 211 At to label monoclonal antibody (BC8, anti-human CD45) for preclinical use, following isolation from the 207 Po contamination also produced by these generators (by 211 Rn α-decay)., Methods: 211 Rn was produced by 211 Fr electron capture decay following mass separated ion beam implantation and chemically isolated in liquid alkane hydrocarbon (dodecane). 211 At produced by the resulting 211 Rn source was extracted in strong base (2N NaOH) and purified by granular Te columns. BC8-B10 (antibody conjugated with closo-decaborate(2-)) was labeled with generator-produced 211 At and purified by PD-10 columns., Results: Aqueous solutions extracted from the generator were found to contain 211 At and 207 Po, isolated from 211 Rn. High radionuclidic purity was obtained for 211 At eluted from Te columns, from which BC8-B10 monoclonal antibody was successfully labeled. If not removed, 207 Po was found to significantly contaminate the final 211 At-BC8-B10 product. High yield efficiencies (decay-corrected, n=3) were achieved for 211 At extraction from the generator (86%±7%), Te column purification (70%±10%), and antibody labeling (76%±2%)., Conclusions: The experimental 211 Rn/ 211 At generator was shown to be well-suited for preclinical 211 At-based research., Advances in Knowledge: We believe that these experiments have furthered the knowledge-base for expanding accessibility to 211 At using the 211 Rn/ 211 At generator system., Implications for Patient Care: As established by this work, the 211 Rn/ 211 At generator has the capability of facilitating preclinical evaluations of 211 At-based therapies., (Copyright © 2017 Elsevier Inc. All rights reserved.)

Copper-64 is a useful radioisotope for positron emission tomography (PET). Due to the wide range of applications, the demand of 64 Cu with low metallic impurities is increasing. Here we report a simple method for the efficient production of high specific activity 64 Cu using a cyclotron for biomedical application. We designed new equipment based on the plating of enriched 64 Ni as the target, and used automated ion exchange chromatography to purify copper-64 efficiently after irradiation and dissolution of the target in good radiochemical and chemical yield and purity. The 64 Cu radionuclide produced using 99.32% enriched 64 Ni with a density of 61.4 ± 5.0 mg/cm², reaching a total radioactivity greater than 200 mCi, with specific activity up to 5.6 GBq/μmoL. It was further incorporated into modified monoclonal antibody DOTA-rituximab to synthesize 64 Cu-DOTA-rituximab, which was used successfully for micro-PET imaging.

The possibility of 123 I isotope production with the help of the high-intensity bremsstrahlung photons produced by the electron beam of the LUE50 linear electron accelerator at the A.I. Alikhanyan National Science Laboratory (Yerevan Physics Institute [YerPhI]) is considered. The production method has been established and shown to be successful. The 124 Xe(γ,n) 123 Xe→ 123 I nuclear reaction has been investigated and the cross-section was calculated by nuclear codes TALYS 1.6 and EMPIRE 3.2. The optimum parameter of the thickness of the target was determined by GEANT4 code. For the normalized yield of 123 I, the value of 143Bq/(mg·μA·h) has been achieved., (Copyright © 2016 Elsevier Inc. All rights reserved.)