Radionuclidic Purity Research Articles

Hydroxyapatite (Ha) labeling with a phosphorus-32 radioisotope of the TRIGA 2000 reactor irradiation result as a candidate for radiosinovectomy therapy

Radiosynovectomy is a therapy performed on patients with acute-level arthritis (rheumatoid arthritis) as an alternative solution besides surgery. Radiosynovectomy is performed using a labeled compound with a particle size of 0.5-10 μm labeled with a β radioisotope. Hydroxyapatite (HA) is a 1-10 μm-sized compound found in bones with the components of Calcium (Ca) & Phosphorus (P). Phosphorus-32 (32P) is a radioactive form of Phosphorus which emits pure beta rays and is often used for therapy. Labelling HA with 32P tends to be easy to do with a substitution reaction, because phosphorus is the main constituent of HA. Phosphorus-32 was made by irradiating natural sulfur at the Bandung TRIGA 2000 reactor facility following the 32S (n, p) 32P reaction mechanism. The separation process of Phosphorus-32 was carried out by a distillation method followed by extraction with 0.01 N HCl accompanied by heating for 30 minutes. The Phosphorus-32 solution is then passed through a 3 gr cation exchange resin. Before Phosphorus-32 was used for Labelling of HA, a Radionuclide Purity test was performed with a gamma-MCA spectrophotometer and a Radiochemical Purity test using paper chromatography. The test results showed Phosphorus-32 had Radionuclide Purity > 99.99% and Radiochemical Purity > 96%. 0.5 mCi Phosphorus-32 which meets the quality test requirements is reacted with 7 mg Ha at pH 7. Then it is vortexed at 1500 rpm for 60 minutes with 70 ° C heating. HA-32P is separated using centrifugation into residual and supernatant fractions. Measure the radioactivity of both fractions with a dose calibrator. Labeling Yield HA with Phosphorus-32 was obtained 98%. Furthermore this HA is ready to be used in in vivo tests for radiosynovectomy.

Validation of Quality Control Parameters of Cassette-Based Gallium-68-DOTA-Tyr3-Octreotate Synthesis.

Purpose of the Study:Gallium (Ga)-68-DOTA peptides targeting somatostatin receptors have been assessed as a valuable tool in neuroendocrine tumor imaging using positron emission tomography. However, at the moment, a specific monograph in the European Pharmacopoeia (Ph. Eur.) does exist only for Ga-68-edotreotide (DOTATOC) injection. Here, we report on the validation process of Ga-68-DOTA-Tyr3-octreotate (DOTATATE) cassette-based production and quality control (QC).Materials and Methods:Preparation of Ga-68-DOTATATE was performed according to the current European Union-good manufacturing practices, the current good radiopharmacy practice, the Ph. Eur., and the guidelines on validation of analytical methods for radiopharmaceuticals. Process was validated via three consecutive production runs to ensure that the methods are reproducible and reliable in routine use. The QC tests for Ga-68-DOTATATE were radiochemical purity (RCP – high-pressure liquid chromatography [HPLC]), radiochemical impurities 68Ga3+ (HPLC and instant thin layer chromatography [ITLC]), chemical purity (HPLC and gas chromatography [GC]), pH (pH-strips), radionuclidic purity (principal γ-photon), germanium-breakthrough (68Ge-content), Ga-68 half-life (γ-ray spectrometry), and sterility/endotoxin assay.Results:Radiolabeling procedure of Ga-68-DOTATATE fits all the applicable Ph. Eur. specifications. RCP measured via ITLC was >99% in the three validation batches. HPLC-measured RCP resulted 99.45%, 99.78%, and 99.75%. Germanium-breakthrough was far below the recommended level established in the Ph. Eur. Ga-68-DOTATOC injection (#2482). Residual ethanol tested with GC was less than 10%. All the batches were tested for endotoxin content, which always resulted lower than 17.5 EU/ml. All preparations passed the sterility tests. pH of the final product was 7 in all samples.Conclusion:Ga-68-DOTATATE fulfilled all the pre-set QCs and release criteria in the batches considered for this validation study. The results demonstrated a batch-to-batch reproducibility, ensuring that synthesis process leads to the expected final product in terms of yield, quality, reliability, safety, and efficacy.

Taking cyclotron 68Ga production to the next level: Expeditious solid target production of 68Ga for preparation of radiotracers

IntroductionGallium-68 is an important radionuclide for positron emission tomography (PET) with steadily increasing applications of 68Ga-based radiopharmaceuticals for clinical use. Current 68Ga sources are primarily 68Ge/68Ga-generators, along with successful attempts of 68Ga production using a cyclotron. This study evaluated cyclotron 68Ga production and automated separation using expeditiously manufactured solid targets, demonstrates an order of magnitude improvement in yield compared to 68Ge/68Ga 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 68Ga. Methods100 mg enriched 68Zn (99.3% 68Zn, 0.48% 67Zn, 0.1% 66Zn) pellets pressed on silver discs were bombarded for 20–75 min using 12.5 MeV proton beam energies and 10–30 μA currents. 68Ga was separated using an automated TRASIS AllinOne synthesizer employing AG 50W-X8 and UTEVA resins. Post-separation recovery of the 68Zn by electrolysis yielded 76.7 ± 4.3%. Radionuclidic purity of cyclotron-produced 68Ga 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 [68Ga]Ga-NOTA-BBN2 in a PC3 xenograft model. Results600 μA·min fresh and recycled quadruplet 68Zn target irradiations (n = 8) at 12.5 MeV and 30 μA yielded 13.9 ± 1.0 GBq 68Ga; 2200 μA·min irradiations (n = 3) yielded 37.5 ± 1.9 GBq 68Ga. HPGe analysis showed EOB 0.0074% and 0.0084% of total activity of 66Ga and 67Ga, 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 68Ga. Cyclotron and 68Ge/68Ga generator 68Ga 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 68Ga showed identical radiotracer tumor uptake and biodistribution profiles in PC3 tumor bearing mice. ConclusionsCyclotron 68Ga production provides highly scalable production with equivalent or superior quality 68Ga to a 68Ge/68Ga 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 68Ga. The manufacturing simplicity described has potential applications for producing other radiometals such as 44Sc. Advances in knowledge and implications for patient careOur cost-effective method of solid target 68Ga production can enhance 68Ga production capabilities to meet the high demand for 68Ga-radiopharmaceuticals for research and clinical use.

Automated radiosynthesis of 68Ga for large-scale routine production using 68Zn pressed target

Gallium-68 (68Ga) has attracted increasing interest in recent years due to the expanding clinical applications of 68Ga-based radiopharmaceuticals (Rahbar et al., 2017). 68Ga is mainly produced via 68Ge/68Ga generators that are limited in yield by the 68Ge activity (typically up to 1.85 GBq at calibration time). With the increased-demand of 68Ga in nuclear medicine for positron emission tomography (PET) imaging, there is a need for more efficient and robust production methods to obtain larger amounts of [68Ga]GaCl3 with high radionuclidic and radiochemical purity and apparent molar activity (AMA) for facilitating the distribution of 68Ga-based radiopharmaceuticals. The objectives of this study were to develop a fast and efficient process for the preparation of 68Zn-based solid targets and to optimize the critical parameters for the automated radiosynthesis of [68Ga]GaCl3 for large-scale routine production.

Production of 230Pa from Proton-irradiated Thorium and Developing 230Pa/230U/226Th Tandem Generator

Targeted alpha-therapy of various oncology diseases is based on the coupling of alpha-emitting radionuclides to tumour-selective carrier molecules. 230U (t1/2 = 20.8 d) is accumulated via decay of 230Pa and has strong potential for alpha-therapy due to 5 emitting alpha particles with total energy 33.5 MeV. It can be used directly or as a parent of 226Th in a generator system. A prospective way for production of 230Pa/ 230U is irradiation of natural thorium with medium-energy protons. Curie amounts of 230Pa can be generated in one irradiation run together with other useful alpha-emitters 225Ac and 223Ra. After dissolving Th-target in 6M nitric acid with the addition of catalytic amounts of hydrofluoric acid Pa was isolated by liquid extraction with octanol. The water phase may be further used for isolation of 225Ac and 223Ra according the method, proposed in [1, 2]. Namely, the most part of thorium was removed by liquid extraction with di(2-ethylhexyl)phosphoric acid in toluene. Organic phase was kept for accumulation 223Ra from parent 227Th. 225Ac and 223Ra were concentrated from aqueous phase and then separated and purified by extraction chromatography. After re-extraction 230Pa was purified on silica gel using different oxalic acid solutions. Up to 85% of 230Pa with radionuclidic purity >99% was recovered and stored for 230U accumulation. Basing on distribution coefficients 230U was separated from 230Pa in diluted nitric acid solution on DGA resin. Obtained 230U may be used for development of 230U/226Th – generator. Distribution coefficients for U and Th were obtained for a wide range of extraction chromatographic resins. TEVA (mixture of trioctyl and tridecyl methyl ammonium chloride as an extractant) and WBEC (based on a mixture of tertiary octyl and decylamines) resins are most suitable for this purpose. [1]R.A. Aliev, S.V. Ermolaev, A.N. Vasiliev, V.S. Ostapenko, E.V. Lapshina, B.L. Zhuikov, N.V. Zakharov, V.V. Pozdeev, V.M. Kokhanyuk, B.F. Myasoedov, S.N. Kalmykov. Isolation of medicine-applicable actinium-225 from thorium targets irradiated by medium-energy protons. Solvent Extraction and Ion Exchange, 2014, v. 32, p. 468-477.[2]A.N. Vasiliev, V.S. Ostapenko , E.V. Lapshina, S.V. Ermolaev, S.S. Danilov, B.L. Zhuikov and S. N. Kalmykov. Recovery of Ra-223 from natural thorium irradiated by protons. Radiochimica Acta, 2016, v. 104, N.8, p. 539-547.

Capabilities of JSC “SSC RIAR” in Producing 227ThCl4

227Th is considered to be one of the promising radionuclides for target radionuclide therapy of cancer. Its advantages are chemical properties which are good for the synthesis of the complexes with chelating agents and emission of five alpha particles with complete decay of one 227Th nucleus. The apparent disadvantage is the presence of 223Ra among daughter products the chemical properties of which are fundamentally different from thorium that can cause its redistribution in the human body. A rather big half-life of 227Th (18.7 days), on the one hand, is good for the preparation production and shipment, and on the other hand, it specifies high-level requirements to the preparation stability in the human body. In addition to using 227Th for radiopharmaceutical synthesis, it can be applied in manufacturing 223Ra generators. While such generators have a rather small period of use, they have a number of advantages over 227Ac/223Ra generators. In particular, there is no need to check every batch for the 227Ac content, and there are no long-lived radioactive wastes. JSC “SSC RIAR” performs experiments to produce trial samples of 227Th by generating from long-lived parent radionuclide 227Ac extracted from the 226Ra targets irradiated during 20-25 days in the SM reactor neutron trap. To separate 227Th and 227, thorium is sorbed on BioRad AG 18 (NO3-) strongly basic anion-exchange resin from 8M HNO3 with further elution using 0.1-0.5 HNO3 or HCl. To produce the preparation of desired radionuclide purity, the purification process must be carried out at least twice. The key challenge in producing 227Th is determination of the 227Ac content. It is not possible to directly determine 227Ac by its own alpha radiation at its activity of less than 1% of the 227Th activity. The yield of alpha radiation in 227Ac decay is only 1.38 %, and its peak in the spectrum is on the low-energy tail of 227Th and its daughter products. 227Th conversion electrons and beta radiation of the daughter products, namely 211Pb and 211Bi, are obstacles in measuring the 227Ac activity by beta radiation. The presentation discusses possible methods to check the content of 227Ac with its preliminary chemical extraction from a preparation aliquot and gives the characteristics of 227Th experimental samples.

Improved production of 76Br, 77Br and 80mBr via CoSe cyclotron targets and vertical dry distillation

IntroductionThe 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 76Br, 77Br, and 80mBr. The radiochemical quality of the radiobromine produced using these methods was tested through the synthesis of a novel 77Br-labeled inhibitor of poly (ADP-ribose) polymerase-1 (PARP-1), a DNA damage response protein. Methods76Br, 77Br, and 80mBr were produced in high radionuclidic purity via the proton irradiation of novel isotopically-enriched Co76Se, Co77Se, and Co80Se intermetallic targets, respectively. Radiobromine was isolated through thermal chromatographic distillation in a vertical furnace assembly. The 77Br-labeled PARP inhibitor was synthesized via copper-mediated aryl boronic ester radiobromination. ResultsCyclotron production yields were 103 ± 10 MBq∙μA−1∙h−1 for 76Br, 88 ± 10 MBq∙μA−1∙h−1 for 80mBr at 16 MeV and 17 ± 1 MBq∙μA−1∙h−1 for 77Br at 13 MeV. Radiobromide isolation yields were 76 ± 11% in a small volume of aqueous solution. The synthesized 77Br-labeled PARP-1 inhibitor had a measured apparent molar activity up to 700 GBq/μmol at end of synthesis. ConclusionsA novel selenium alloy target enabled clinical-scale production of 76Br, 77Br, and 80mBr with high apparent molar activities, which was used to for the production of a new 77Br-labeled inhibitor of PARP-1. Advances in knowledgeNew methods for the cyclotron production and isolation of radiobromine improved the production capacity of 77Br by a factor of three and 76Br by a factor of six compared with previous methods. Implications for patient carePreclinical translational research of 77Br-based Auger electron radiotherapeutics, such as those targeting PARP-1, will require the production of GBq-scale 77Br, which necessitates next-generation, high-yielding, isotopically-enriched cyclotron targets, such as the novel intermetallic Co77Se.

Electromagnetic isotope separation of gadolinium isotopes for the production of 152,155Tb for radiopharmaceutical applications

A demo experiment was performed at the tandem accelerator of MLL Garching where 152Tb was produced by irradiating a unique ion-implanted 152Gd target (>99% enriched) with 8 and 12 MeV protons respectively. At these energies mainly 152Tb was produced by (p,n) reactions plus a small quantity of 153Tb by (p,γ) reactions. Upper limits are derived for co-production of other Tb isotopes. The achieved radionuclidic purity would enable direct use for human applications. Gd isotopes of adequate isotopic enrichment could be separated with the SIDONIE mass separator at CSNSM Orsay and prepared as cyclotron targets suited for high current irradiations. In a demo experiment samples of highly enriched 158Gd were produced at SIDONIE. The sample purity was characterized by prompt gamma activation analysis at NPI Řež. The neighboring mass 157Gd was found to be suppressed by three to five orders of magnitude, depending on the sample position in the SIDONIE separator.

The ISOLPHARM project: ISOL-based production of radionuclides for medical applications

Radionuclides for radiopharmaceuticals can be produced in cyclotrons or nuclear reactors. Each of these production modes has serious issues, such as high target costs, production of long-lived wastes and contaminants, expensive separation. For this reason, new methods are under consideration for the production of highly pure radionuclides. The ISOL (Isotope Separation On-Line) method is the major technique for the production of radioactive ion beams for nuclear physics applications. The SPES-ISOLPHARM project at INFN-LNL (Istituto Nazionale di Fisica Nucleare-Laboratori Nazionali di Legnaro) is a feasibility study for the production of medical isotopes exploiting the ISOL method. The ongoing activities concerning a recent experiment focused on 111Ag, a study performed in collaboration with Padova and Trento Universities, is presented.

Purification of 99Mo and 99mTc from radioactive traces of Nb, Zr, and Y impurities: method applicable in the purification of the spent 100/99Mo–99mTc generator

The aim of this work was to develop a method for purification of the spent 100/99Mo–99mTc generator free from non-isotopic impurities like Nb, Y, Zr and recovery of 99mTc using a Sephadex column and MEK solvent extraction technique, respectively. This study was done by performing simulation experiment with 98Mo(n,γ)99Mo solution doped with the radiotracers of the above said non-isotopic impurities. There was respective adsorption of non-isotopic impurities of Nb, Zr and Y on Sephadex column (99.99%; n = 3) and extraction 99mTc in MEK. The extraction efficiency of pure 99mTc (radionuclidic purity: 99.99%, n = 3) from various Mo solution was 97% (n = 3).

Chemical Purification of Terbium-155 from Pseudo-Isobaric Impurities in a Mass Separated Source Produced at CERN

Four terbium radioisotopes (149, 152, 155, 161Tb) constitute a potential theranostic quartet for cancer treatment but require any derived radiopharmaceutical to be essentially free of impurities. Terbium-155 prepared by proton irradiation and on-line mass separation at the CERN-ISOLDE and CERN-MEDICIS facilities contains radioactive 139Ce16O and also zinc or gold, depending on the catcher foil used. A method using ion-exchange and extraction chromatography resins in two column separation steps has been developed to isolate 155Tb with a chemical yield of ≥95% and radionuclidic purity ≥99.9%. Conversion of terbium into a form suitable for chelation to targeting molecules in diagnostic nuclear medicine is presented. The resulting 155Tb preparations are suitable for the determination of absolute activity, SPECT phantom imaging studies and pre-clinical trials.

Radiometals for imaging and theranostics, current production, and future perspectives.

The aim of this review is to make the reader familiar with currently available radiometals, their production modes, capacities, and quality concerns related to their medical use, as well as new emerging radiometals and irradiation technologies from the perspective of their diagnostic and theranostic applications. Production methods of 177 Lu serve as an example of various issues related to the production yield, specific activity, radionuclidic and chemical purity, and production economy. Other radiometals that are currently used or explored for potential medical applications, with particular focus on their theranostic value, are discussed. Using radiometals for diagnostic imaging and therapy is on the rise. The high demand for radiometals for medical use prompts investigations towards using alternative irradiation reactions, while using existing nuclear reactors and accelerator facilities. This review discusses these production capacities and what is necessary to cover the growing demand for theranostic nuclides.

Production and in vivo PET/CT imaging of the theranostic pair 132/135La

The present study describes a novel method for the low energy cyclotron production and radiochemical isolation of no-carrier-added 132/135La3+ from bulk natBa. This separation strategy combines precipitation and single-column extraction chromatography to afford an overall radiochemical yield (92 ± 2%) and apparent molar activity (22 ± 4 Mbq/nmol) suitable for the radiolabeling of DOTA-conjugated vectors. The produced 132/135La3+ has a radiochemical and radionuclidic purity amenable for 132La/135La-based cancer theranostic applications. Longitudinal PET/CT images acquired using the positron-emitting 132La and ex vivo biodistribution data separately corroborated the accumulation of unchelated 132/135La3+ ions in bone and the liver.

Feasibility study on production of Sc-47 from neutron irradiated Ca target for cancer theranostics applications

Abstract Scandium-47 is one of the most useful radioisotopes which is gaining great importance in cancer theranostics applications due to its favorable nuclear and chemical properties. MCNPX2.7.0 code was used to simulate the neutron activation of natural calcium target positioned at a thermal neutron flux of 1.8 × 1014 n cm−2 s−1 in the Egypt Second Research Reactor (ETRR-2). The burn card was used to calculate 47Ca and 47Sc radioactivities during 3 days irradiation and 20 days post-irradiation. The undesirable impurities generated during this period were also calculated. The obtained calculations were found to be in agreement with the experimental measurements. The distribution coefficient value (Kd) of 47Sc(III) as well as 47Ca(II) ions was determined using the commercially available ion-exchanger Chelex 100 in HNO3 and/or HCl media. Radiochemical separation of 47Sc(III) from 47Ca(II) was studied using HNO3 and HCl solutions and the results showed that HNO3 is a better medium than HCl for complete retention and recovery of 47Sc(III), where the recovery yields were 85 ± 1.2 and 95 ± 0.87 % using 1 M HCl and 1 M HNO3 solutions, respectively. The recovery yield obtained in our work was higher than in the reported procedures. Radionuclidic, radiochemical and chemical purities were investigated to ensure the suitability of 47Sc(III) for nuclear medicine applications.

Production and characterization of no-carrier-added 161Tb as an alternative to the clinically-applied 177Lu for radionuclide therapy

Background161Tb is an interesting radionuclide for cancer treatment, showing similar decay characteristics and chemical behavior to clinically-employed 177Lu. The therapeutic effect of 161Tb, however, may be enhanced due to the co-emission of a larger number of conversion and Auger electrons as compared to 177Lu. The aim of this study was to produce 161Tb from enriched 160Gd targets in quantity and quality sufficient for first application in patients.MethodsNo-carrier-added 161Tb was produced by neutron irradiation of enriched 160Gd targets at nuclear research reactors. The 161Tb purification method was developed with the use of cation exchange (Sykam resin) and extraction chromatography (LN3 resin), respectively. The resultant product (161TbCl3) was characterized and the 161Tb purity compared with commercial 177LuCl3. The purity of the final product (161TbCl3) was analyzed by means of γ-ray spectrometry (radionuclidic purity) and radio TLC (radiochemical purity). The radiolabeling yield of 161Tb-DOTA was assessed over a two-week period post processing in order to observe the quality change of the obtained 161Tb towards future clinical application. To understand how the possible drug products (peptides radiolabeled with 161Tb) vary with time, stability of the clinically-applied somatostatin analogue DOTATOC, radiolabeled with 161Tb, was investigated over a 24-h period. The radiolytic stability experiments were compared to those performed with 177Lu-DOTATOC in order to investigate the possible influence of conversion and Auger electrons of 161Tb on peptide disintegration.ResultsIrradiations of enriched 160Gd targets yielded 6–20 GBq 161Tb. The final product was obtained at an activity concentration of 11–21 MBq/μL with ≥99% radionuclidic and radiochemical purity. The DOTA chelator was radiolabeled with 161Tb or 177Lu at the molar activity deemed useful for clinical application, even at the two-week time point after end of chemical separation. DOTATOC, radiolabeled with either 161Tb or 177Lu, was stable over 24 h in the presence of a stabilizer.ConclusionsIn this study, it was shown that 161Tb can be produced in high activities using different irradiation facilities. The developed method for 161Tb separation from the target material yielded 161TbCl3 in quality suitable for high-specific radiolabeling, relevant for future clinical application.

Preparation and Evaluation of 177Lu-TDTMP: A Potential Theranostic Agent for Bone Metastases

177Lu was produced in high-specific activity and excellent radionuclidic purity by thermal neutron bombardment of 176Yb target.Tetramethylene diamine tetramethylenephosphonic acid (TDTMP) was radiolabeled with 177Lu and complexation parameters were studied and optimized: 8.4 µg TDTMP, ~243 Ci of 177Lu activity, pH= 8 and the reaction mixture was kept at room temperature for 30 min.The complex, formed with high yield (>99%) and retained its stability for more than 14 days. The complex structure was predicted and its binding to target enzyme was assessed in-silico..Biodistribution study carried out in normal and tumorized Wister rats showed significant skeletal uptake, the 177Lu-TDTMP, results clearly indicated that it has high affinity for bone mineral. Femur and tibia uptakes of the labelled complex were ~ 6.4 % and ~7.1 % at 3 hours post injection compared to ~ 4 % and ~ 4.3 % in case of 177LuCl3. Thereafter, the uptakes increased continuously to a maximum of ~ 7.9 % and ~ 8.3 % at 48 hours after injection were noticed.

Implementation of a new separation method to produce qualitatively improved 64 Cu.

64 Cu (T1/2 =12.7h) is an important radionuclide for diagnostic purposes and used for positron emission tomography (PET). A previous method utilized at Paul Scherrer Institute (PSI) proved to be unreliable and, while a method using anion exchange chromatography is a popular choice worldwide, it was felt a different approach was required to obtain a robust chemical separation method. Enriched 64 Ni targets were created by electroplating on gold foil. The targets were irradiated with protons degraded to approximately 11MeV at PSI's Injector 2 72 MeV research cyclotron and subsequently dissolved in HCl. The resultant solution was loaded onto AG MP-50 cation exchange resin and the 64 Cu separated from its target material and radiocobalt impurities, produced as part of the irradiation process, using various specific mixtures of HCl/acetone solution. The eluted product was evaporated and picked up in dilute HCl (0.05M). The chemical purity of 64 Cu was determined by radiolabeling experiments at the highest possible molar activities. Reproducible results were obtained, yielding 3.6 to 8.3 GBq 64 Cu of high radionuclidic and radiochemical purity. The product was labeled to NODAGA-RGD, achieved at up to 500MBq/nmol, indicating the high chemical purity. In a proof-of-concept in vivo study, 64 Cu-NODAGA-RGD was used for PET imaging of a tumor-bearing mouse. The chemical separation devised to produce high-quality 64 Cu proved to be robust and reproducible. The concept can be used at medical cyclotrons utilizing a solid target station, such that 64 Cu can be used at hospitals for PET imaging.

Improved procedures of Sc(OH)3 precipitation and UTEVA extraction for 44Sc separation.

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. 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. 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. 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.

Radiochemical separation of reactor produced Sc-47 from natural calcium target using Poly(acrylamide-acrylic acid)/multi-walled carbon nanotubes composite

Poly (acrylamide-acrylic Acid)/multi-walled carbon nanotubes composite [P(AAm-AA)/MWCNTs] was prepared by gamma radiation to initiate the copolymerization of AAm and AA onto MWCNTs surface. The structural characteristics were confirmed by FT-IR, SEM, XRD, TGA and DTA. The particle size distribution was determined by dynamic light scattering (DLS). [P(AAm-AA)/MWCNTs] was used to study the adsorption efficiency of 47Sc from natural calcium by batch technique at different concentrations of HCl and/or HNO3 acid solutions. No-carrier-added 47Sc (III) was separated from 47Ca(II) on [P(AAm-AA)/MWCNTs] column matrix. The elution efficiency of about 80 ± 1.5% was attained for 47Sc(III) using 1M HNO3. Three quality control tests, radionuclidic, radiochemical and chemical purities were investigated to emphasis that 47Sc secure for use in nuclear medicine.

Production of [13N]ammonia from [13C]methanol on a 7.5 MeV cyclotron using 13C(p, n)13N reaction: Detection of myocardial infarction in a mouse model

[13N]Ammonia is commonly produced using 16O(p, α)13N reaction but one of the limiting factor of this reaction is the relatively small nuclear cross-section at proton energies of <10 MeV. An alternative production method using 13C(p, n)13N reaction, which has a higher nuclear cross-section at low proton energies, is more suitable for a preclinical PET imaging facility equipped with a <10 MeV cyclotron. Here, we report a novel method to produce [13N]ammonia from [13C]methanol for preclinical use on a 7.5 MeV cyclotron. A tantalum solution target (80 μl) consisting of a havar window supplied by the cyclotron manufacturer for the production of [18F]fluoride was used without any modifications. The final bombardment parameters were optimized as follow: [13C]methanol concentration in target solution – 10%, bombardment time – 8 min, and beam current – 2.2 μA. These parameters provided doses of [13N]ammonia which were sufficient to conduct preclinical PET imaging studies in a mouse model of myocardial infarction. Under optimized conditions, the operational lifetime of the target was approximately 150 μAmin. Radionuclide identity of the product as 13N was confirmed by measuring the decay half-life and its radionuclide purity was confirmed by γ-ray spectroscopic analysis. Gas chromatography revealed that the final [13N]ammonia dose was not distinguishable from water, showing no traces of methanol. As expected, PET/CT imaging in healthy CD-1 mice indicated the accumulation of [13N]ammonia in myocardial tissue; mice with myocardial infarction created by left ascending coronary ligation showed clear perfusion deficit in affected tissue. This work demonstrates the proof-of-concept of using 13C(p, n)13N reaction to produce [13N]ammonia from [13C]methanol with a <10 MeV cyclotron, and its diagnostic application in imaging cardiac perfusion.