Preparation of 177Lu-PSMA-617 in Hospital Radiopharmacy: Convenient Formulation of a Clinical Dose Using a Single-Vial Freeze-Dried PSMA-617 Kit Developed In-House.

OBJECTIVE: In the recent time, endoradionuclide therapy for metastatic castration-resistant prostate carcinoma employing 177Lu-PSMA-617 has yielded encouraging results and several clinical trials with the agent are currently ongoing. Routine preparation of 177Lu-PSMA-617 patient doses can be made simpler and convenient, if the ingredients essential for radiolabeling are made available in a ready-to-use lyophilized form. METHODS: PSMA-617 freeze-dried kit was formulated and used for the preparation of 177Lu-PSMA-617 clinical dose with high radiochemical purity using low/medium specific activity 177Lu. Detailed radiochemical studies were performed to determine the maximum activity and volume of 177LuCl3, which can be added in the kit for the formulation of 177Lu-PSMA-617. Studies were also performed to determine the shelf life of the kit to ensure its long-term usage. Studies were performed in buffer as well as human serum medium to determine the stability of the 177Lu-PSMA-617 complex after storing in respective media up to 7 days postpreparation. About ten patient doses of 177Lu-PSMA-617 were administered, and posttherapy scans were acquired. RESULTS: The formulated freeze-dried kit of PSMA-617 could be radiolabeled with an average percentage radiochemical purity > 98.53 ± 0.38. The freeze-dried kit was found suitable for tolerating up to 0.5 mL of 177LuCl3 (in 0.01 N HCl) and specific activity of 555 MBq/µg (15 mCi/µg) for the preparation of the patient dose of 177Lu-PSMA-617. The 177Lu-PSMA-617 complex prepared using the freeze-dried kit of PSMA-617 was observed to maintain % radiochemical purity (RCP) of 96.74 ± 0.87 and 94.81 ± 2.66, respectively, even after storing up to 7 days in buffer and human serum, respectively. 177Lu-PSMA-617 prepared using the in-house formulated freeze-dried kit of PSMA-617 exhibited accumulation in metastatic lesions picked up in a pretherapy PET scan. Reduction in number as well as size of lesions was observed in posttherapy scans acquired after two months of administering the first therapeutic dose of 177Lu-PSMA-617. CONCLUSIONS: The freeze-dried kit of PSMA-617 could be used for the preparation of 177Lu-PSMA-617 with high radiochemical purity (>98%) in a reproducible manner. 177Lu-PSMA-617 prepared using the developed kit was successfully evaluated in patients suffering from metastatic prostate cancer.

Green tea essential oil encapsulated chitosan nanoparticles-based radiopharmaceutical as a new trend for solid tumor theranosis.

The existing study is embarked on investigating the antineoplastic activity of green tea essential oil (GTO) as a natural product. In this regard, GTO was encapsulated in cationic chitosan, nitrogenous-polysaccharide derived by partial deacetylation of chitin, nanoparticles (CS NPs) with entrapment efficiency (EE%) of 81.4 ± 5.7% and a mean particle-size of 30.7 ± 1.13 nm. Moreover, the cytotoxic effect of CS/GTO NPs was evaluated versus human liver (HepG-2), breast (MCF-7) and colon (HCT-116) cancer cell-lines and exhibited a positive impact when compared to bare CS NPs by 3, 2.3 and 1.7 fold for the three cell lines, respectively. More interestingly, CS/GTO NPs were complexed with technethium-99m (99mTc) radionuclide. With a view to achieve a successful radiolabeling process, different parameters were optimized resulting in a radiolabeling efficiency (RE%) of 93.4 ± 1.2%. Radiopharmacokinetics of the radiolabeled NPs in healthy mice demonstrated a reticuloendothelial system (RES) evading and long blood circulation time up to 4 h. On the other hand, the biodistribution profile in solid tumor models showed 20.3 ± 2.1% localization and cancer cell targeting within just 30 min. On the whole, the reported results encourage the potential use of CS/GTO NPs as a side effect-free anticancer agent and its 99mTc-analogue as a novel CS/GTO NPs-based diagnostic-radiopharmaceutical for cancer.

Application of chiral chromatography in radiopharmaceutical fields: A review.

Chiral column chromatography (CCC) is a revolutionary analytical methodology for the enantioseparation of novel positron emission tomography (PET) tracers in the primary stages of drug development. Due to the different behaviors of tracer enantiomers (e.g. toxicity, metabolism and side effects) in administrated subjects, their separation and purification is a challenging endeavor. Over the last three decades, different commercial chiral columns have been applied for the enantioseparation of PET-radioligand (PET-RL) or radiotracers (PET-RT), using high-performance liquid chromatography (HPLC). The categorization and reviewing of them is a vital topic. This review presents a brief overview of advances, applications, and future prospectives of CCC in radiopharmaceutical approaches. In addition, the effective chromatographic parameters and degravitation trends to enhance enantioseparation resolution are addressed. Moreover, the application and potential of chiral super fluidical chromatography (CSFC) as an alternative for enantioseparation in the field of radiopharmaceutical is discussed. Finally, the crucial application challenges of CCC are explained and imminent tasks are suggested.

Biosynthetic new composite material containing CuO nanoparticles produced by Aspergillus terreus for 47Sc separation of cancer theranostics application from irradiated Ca target.

Copper oxide nanoparticles (CuO NPs) are needed in various fields, especially in the biomedical field. CuO NPs was obtained from Aspergillus terreus filtrate. CuO NPs structure was confirmed by UV-Vis spectrophotometry, as well as Fourier transform infrared (FT-IR), X-ray diffraction (XRD), scanning electron microscopy (SEM) and transmission electron microscopy (TEM). XRD offers the nanoparticles purity of CuO biosynthesis. CuO NPs are spherical when examined with TEM. The average size of CuO NPs from TEM was 15.75 ± 3.95 nm. New composite of P (AA-AN)-NPs CuO was synthesized by biotechnology and the induced γ-radiation. The distribution coefficient value (Kd) of 47Sc(III) as well as 47Ca(II) ions for the synthetic new composite was determined by batch technique. Radiochemical separation of 47Sc(III) from irradiated calcium target was studied using chromatographic column packed with the new composite material. The recovery yield of 78 ± 1.2% for 47Sc(III) was obtained using 1 M HCl. The quality control tests (chemical, radionuclide and radiochemical purities) of the eluted 47Sc confirmed that it's adequate for nuclear medicine applications.

A simple and robust method for radiochemical separation of no-carrier-added 64Cu produced in a research reactor for radiopharmaceutical preparation.

Copper-64 is an excellent theranostic radiometal that is gaining renewed attention of the clinical community in the recent times. In order to meet the increasing demand of this radiometal, we have demonstrated the viability of its production via 64Zn (n,p) 64Cu reaction in a nuclear reactor. A semi-automated radiochemical separation module based on selective extraction of 64Cu as dithizonate complex was developed. The maximum available activity at the end of irradiation was ~ 700 MBq. The overall yield of 64Cu after the separation process was >85% and it could be obtained with ~12 GBq/µg specific activity, >99.9% radionuclidic purity and >98% radiochemical purity. The separated 64Cu could be utilized for preparation of a wide variety of radiopharmaceuticals.

Production and Semi-Automated Processing of 89Zr Using a Commercially Available TRASIS MiniAiO Module.

The increased interest in 89Zr-labelled immunoPET imaging probes for use in preclinical and clinical studies has led to a rising demand for the isotope. The highly penetrating 511 and 909 keV photons emitted by 89Zr deliver an undesirably high radiation dose, which makes it difficult to produce large amounts manually. Additionally, there is a growing demand for Good Manufacturing Practices (GMP)-grade radionuclides for clinical applications. In this study, we have adopted the commercially available TRASIS mini AllinOne (miniAiO) automated synthesis unit to achieve efficient and reproducible batches of 89Zr. This automated module is used for the target dissolution and separation of 89Zr from the yttrium target material. The 89Zr is eluted with a very small volume of oxalic acid (1.5 mL) directly over the sterile filter into the final vial. Using this sophisticated automated purification method, we obtained satisfactory amount of 89Zr in high radionuclidic and radiochemical purities in excess of 99.99%. The specific activity of three production batches were calculated and was found to be in the range of 1351-2323 MBq/µmol. ICP-MS analysis of final solutions showed impurity levels always below 1 ppm.

Realizing Clinical Trials with Astatine-211: The Chemistry Infrastructure.

Despite the consensus around the clinical potential of the α-emitting radionuclide astatine-211 (211At), there are only a limited number of research facilities that work with this nuclide. There are three main reasons for this: (1) Scarce availability of the nuclide. Despite a relatively large number of globally existing cyclotrons capable of producing 211At, few cyclotron facilities produce the nuclide on a regular basis. (2) Lack of a chemical infrastructure, that is, isolation of 211At from irradiated targets and the subsequent synthesis of an astatinated product. At present, the research groups that work with 211At depend on custom systems for recovering 211At from the irradiated targets. Setting up and implementing such custom units require long lead times to provide a proper working system. (3) The chemistry of 211At. Compared with radiometals there are no well-established and generally accepted synthesis methods for forming sufficiently stable bonds between 211At and the tumor-specific vector to allow for systemic applications. Herein we present an overview of the infrastructure of producing 211At radiopharmaceuticals, from target to radiolabeled product including chemical strategies to overcome hurdles for advancement into clinical trials with 211At.

Rapid and automated production of [68Ga]gallium chloride and [68Ga]Ga-DOTA-TATE on a medical cyclotron.

INTRODUCTION: The demand for Gallium-68 (68Ga) for labelling PET radiopharmaceuticals has increased over the past few years. 68Ga is obtained through the decayed parent radionuclide 68Ge using commercial 68Ge/68Ga generators. The principal limitation of commercial 68Ge/68Ga generators is that only a limited and finite quantity of 68Ga (<1.85 GBq at start of synthesis) may be accessed. The focus of this study was to investigate the use of a low energy medical cyclotron for the production of greater quantities of 68Ga and to develop an automated and rapid procedure for processing the product. METHODS: Enriched ZnCl2 was electrodeposited on a platinum backing using a NH4Cl (pH 2-4) buffer. The Zn target was irradiated with GE PETtrace 880 at 35 µA and 14.5 and 12.0 MeV beam energy. The irradiated Zn target was purified using octanol resin on an automated system. RESULTS: Following the described procedure, 68Ga was obtained in 6.30 ±â€¯0.42 GBq after 8.5 min bombardment and with low radionuclidic impurities (66Ga (<0.005%) and 67Ga (<0.09%)). Purification on a single octanol resin gave 82% recovery with resulting [68Ga]GaCl3 obtained in 3.5 mL of 0.2 M HCl. [68Ga]GaCl3 production from irradiation to final product was <45 min. To highlight the utility of the automated procedure, [68Ga]Ga-DOTA-TATE labelling was incorporated to give 1.56 GBq at EOS of the labelled peptide with RCY of >70%. CONCLUSIONS: A straightforward procedure for producing 68Ga on a low energy medical cyclotron was described. Current efforts are focus on high activity production and radiolabelling using solid target produced 68Ga.

VMAT2 imaging agent, D6-[18F]FP-(+)-DTBZ: Improved radiosynthesis, purification by solid-phase extraction and characterization.

OBJECTIVES: Recently, a deuterated tracer, D6-[18F]FP-(+)-DTBZ, 9-O-hexadeutero-3-[18F]fluoropropoxyl-(+)-dihydrotetrabenazine ([18F]9), targeting vesicular monoamine transporter 2 (VMAT2) in the central nervous system, was reported as a useful imaging agent for the diagnosis of Parkinson's disease (PD). The production of [18F]9 was optimized and simplified by using solid-phase extraction (SPE) purification. METHODS: Three major nonradioactive impurities were synthesized and characterized. The preparation of [18F]9 was optimized by using different labeling conditions, and an SPE purification method was evaluated. The influence of chemical impurities in the final dose of [18F]9 was assessed by an in vitro binding assay, an assay of the in vivo biodistribution in mice, and ex vivo and in vitro autoradiography of brain sections. RESULTS: Optimized fluorination conditions for [18F]9 were found - heating at 130 °C for 10 min in DMSO, and a high radiochemical yield and three major chemical impurities were observed. An SPE method involving a Sep-Pak® tC18 Plus Light cartridge with a two-step elution process was successfully implemented. This process gave a good radiochemical yield (38.7 ±â€¯10.5%, decay corrected; radiochemical purity >99%) and low chemical impurities. An in vivo biodistribution study and autoradiography of brain sections showed that there was no significant difference between HPLC-purified and SPE-purified [18F]9. CONCLUSION: A VMAT2 targeting imaging agent, D6-[18F]FP-(+)-DTBZ, [18F]9, was prepared by optimized labeling conditions and an easy SPE purification. This method offers a short preparation time and operational simplicity. In conjunction with PET imaging, this new VMAT2 agent might be a useful clinical tool for diagnosing PD.

An improved automated one-pot synthesis of O-(2-[18F]fluoroethyl)-L-tyrosine ([18F]FET) based on a purification by cartridges.

INTRODUCTION: O-(2-[18F]Fluoroethyl)-L-tyrosine ([18F]FET) is an established radiotracer used for oncology investigations by Positron Emission Tomography (PET). Main limitations to its widespread use are the synthesis itself (time; cost; radiochemical yield; complexity) and a troublesome and time-consuming HPLC purification. Aim of this work was to improve the preparation overall efficiency and, most important, to achieve an efficient and reliable purification by means of disposable cartridges. METHODS: [18F]FET was synthesized by direct nucleophilic radiofluorination of O-(2-tosyloxy-ethyl)-N-trityl-L-tyrosine t-butylester (TET) followed by acid hydrolysis with HCl. Several conditions and materials were tested for the synthesis and purification step. For the latter, a number of different commercial cartridges, varying in amount, particulate size and adsorbent, were examined. Best results were obtained by a combination of STRATA-X, tC18 and QMA cartridges. RESULTS: Starting from only 5 mg of TET, up to 11 GBq of injectable solutions of [18F]FET were produced within 36 min with 54-65% radiochemical yields and radiochemical purities >99%. No D-form was observed by chiral HPLC. Chemical purity was 1-2 order of magnitude below the limits imposed by the European Pharmacopoeia's monograph on [18F]FET. A radiochemical purity decrease by radiolysis, observed only on relatively large batches of [18F]FET, was efficiently suppressed by preloading in the receiving final vial a small amount of ethanol (<2% v/v). CONCLUSIONS: By combining improvements to a known synthetic route with a novel cartridge-based purification, [18F]FET was obtained in a very efficient and reproducible way. The whole process was easily implemented on a commercial automated module presently used for [18F]FDG production. ADVANCES IN KNOWLEDGE AND IMPLICATIONS FOR PATIENT CARE: A few drawbacks regarding the HPLC conditions recommended in the European Pharmacopoeia were highlighted. An alternative method able to cope with them is herein proposed The simplified preparation herein described is expected to encourage a more widespread clinical use of [18F]FET.

Cyclotron production of 68Ga in a liquid target: Effects of solution composition and irradiation parameters.

OBJECTIVES: To optimize 68Ga production using a liquid cyclotron target, investigations were performed to compare production yields using different concentrations of [68Zn]Zn(NO3)2, nitric acid, and irradiation parameters. METHODS: Different concentrations of [68Zn]Zn(NO3)2 (0.6 M, 1.2 M and 1.42 M) in varying normality of nitric acid (0.8-1.5 N) were prepared and irradiated with protons (incident energy ~14 MeV) using a BMLT-2 liquid target at different beam currents (30-50 µA) and irradiation times (30-60 min). The 68Ga production and saturation yields were calculated and compared. [68Ga]GaCl3 was isolated using in-house developed hydroxamate resin and optimized for routine application. Recycling of [68Zn]Zn(NO3)2 from the recovered target solution was also investigated. RESULTS: On increasing concentration of [68Zn]Zn(NO3)2 from 0.6 M to 1.2 M in 0.8 N nitric acid, decay corrected yield of 68Ga at EOB was found to be 1.64 GBq (44.4 mCi) and 3.37 GBq (91.0 mCi), respectively at 30 µA beam current, indicating production yield was proportional to zinc nitrate concentration for a 30 min irradiation. However, when beam current was increased to 40 µA while maintaining nitric acid concentration at 0.8 N, the proportional relationship of 68Zn-concentration with 68Ga production yield was lost [0.6 M, 2.29 GBq (61.9 mCi); 1.2 M, 3.6 GBq (97.3 mCi)] for a 30 min irradiation. In fact, the effect was more profound for 60 min irradiations [0.6 M, 2.96 GBq (80.0 mCi); 1.2 M, 4.25 GBq (115 mCi)]. Increasing nitric acid concentration to 1.25-1.5 N improved 68Ga production yield for 40 µA, 60-min irradiations (1.2 M; 5.17 GBq (140 mCi)). MP-AES analysis showed metal impurities as <0.20 µg Ga (n = 3), <0.93 µg Zn (n = 3) and < 2.7 µg Fe (n = 3). Based on above finding, 1.42 M [68Zn]Zn(NO3)2 in 1.2 N-HNO3 solutions were also studied to achieve highest production yields of 9.85 ±â€¯2.09 GBq (266 ±â€¯57 mCi) for 60 min irradiation at 40 µA beam current. After recycling,> 99% pure recycled [68Zn]zinc nitrate was obtained in 82.6 ±â€¯13.6% yield. CONCLUSIONS: 68Ga production yields were dependent on all four variables: concentrations of [68Zn]Zn(NO3)2 and nitric acid, beam current and duration of irradiation. Of note, increasing beam current and irradiation time may require increased concentrations of nitric acid to achieve expected increments in 68Ga production yield.

Radiosynthesis of a carbon-11-labeled AMPAR allosteric modulator as a new PET radioligand candidate for imaging of Alzheimer's disease.

To develop PET tracers for imaging of Alzheimer's disease, a new carbon-11-labeled AMPAR allosteric modulator 4-cyclopropyl-7-(3-[11C]methoxyphenoxy)-3,4-dihydro-2H-benzo[e][1,2,4]thiadiazine 1,1-dioxide ([11C]8) has been synthesized. The reference standard 4-cyclopropyl-7-(3-methoxyphenoxy)-3,4-dihydro-2H-benzo[e][1,2,4]thiadiazine 1,1-dioxide (8) and its corresponding desmethylated precursor 4-cyclopropyl-7-(3-hydroxyphenoxy)-3,4-dihydro-2H-benzo[e][1,2,4]thiadiazine 1,1-dioxide (9) were synthesized from 4-methoxyabiline and chlorosulfonyl isocyanate in eight and nine steps with 3% and 1% overall chemical yield, respectively. The target tracer [11C]8 was prepared from the precursor 9 with [11C]CH3OTf through O-[11C]methylation and isolated by HPLC combined with SPE in 10-15% radiochemical yield, based on [11C]CO2 and decay corrected to end of bombardment (EOB). The radiochemical purity was >99%, and the molar activity (AM) at EOB was 370-740 GBq/µmol with a total synthesis time of 35-40-minutes from EOB.

Development of a remote purification apparatus with disposable evaporator for the routine production of high-quality 64Cu for clinical use.

We developed a new apparatus for the routine production of 64Cu in clinical use. The apparatus has many disposable parts that stabilize the product quality (such that there is a low deviation of the concentrations of impurity metals in the product) and reduce the work load of preparation for routine production. We also developed a new evaporator using near-infrared heaters for disposable use. We conducted a production test using the new apparatus and evaluated product quality. The product yield was 6.3 ±â€¯0.32 GBq (end of bombardment) (N = 4), the product quality in terms of the concentrations of impurity metals (Cu2+, Ni2+, Fe3+, Zn2+, Mn2+) was as good as that usually achieved, likely on the order of parts per billion, and the preparation time was reduced from 2 days to 1 day.

Measurement of the 43Sc production cross-section with a deuteron beam.

43,44Sc/47Sc is one of the most promising theranostic pairs in nuclear medicine. The co-emission of 1157 keV γ-rays with 99.9% branching ratio by 44Sc and the presence of its metastable state 44 mSc push to favour the adoption of 43Sc for Positron Emission Tomography (PET) diagnostic procedures to lighten the dose to the patient and to the personnel. The ß+ emitter 43Sc can be produced at a medical cyclotron by proton bombardment of an enriched 43Ca or 46Ti oxide target. 43Sc can be also produced by deuteron bombardment of an enriched 42Ca oxide target. Only a few medical cyclotrons currently in operation offer deuteron beams. Some can be adapted to operate both a proton or a deuteron source. To compare these three production routes, an accurate knowledge of the cross-sections is essential. In this paper, we report on the cross-section measurement of the reaction 42Ca(d,n)43Sc performed at the 6 MV HVEC EN-Tandem of the Ion Beam Physics group at ETH in Zürich. A study of the production yield by using commercially available enriched target materials is also presented.

Studies on batch formulation of a freeze dried kit for the preparation of 99mTc-HYNIC-TATE for imaging neuroendocrine tumors.

AIM: To formulate freeze dried cold kits for preparation of 99mTc-HYNIC-TATE suitable for use at hospital radiopharmacy and establish clinical utility of 99mTc-HYNIC-TATE prepared using kits for detection of neuroendocrine tumors (NETs). METHODS: Standardization of reagent concentrations for formulation of freeze dried kits of HYNIC-TATE was carried out. Consistency in formulation was tested by six batch preparation. Quality control tests were carried out to establish compliance of specifications of purity and safety criteria for both kits and 99mTc-HYNIC-TATE formulated using kits. Clinical utility of 99mTc-HYNIC-TATE prepared using kits was demonstrated in patients with histopathologically confirmed well-differentiated NETs. RESULTS: Pharmaceutical grade HYNIC-TATE kits compliant with all the quality control criteria were formulated and successfully radiolabeled with 99mTc. Radiopharmaceutical was successfully utilized for detection of NETs in patients and comparison with uptake of 99mTc-HYNIC-TOC and 177Lu-DOTA-TATE was made. CONCLUSION: The formulated kits are robust and provide consistently high radiolabeling yields (> 95%) with 99mTc in short time periods requiring no additional purification. Initial clinical trials demonstrate the utility of 99mTc-HYNIC-TATE using formulated kits.

Experimental and theoretical study of rhenium radioisotopes production for manufacturing of new compositional radiopharmaceuticals.

Rhenium therapeutic radioisotopes, namely rhenium-186 and 188, are radionuclides that have been used in combination with various ligands to provide different radiopharmaceuticals for the treatment of different diseases for many years. Each of these radioisotopes has its own special attributes, which make it appropriate to destroy special-sized tumors. High energy, long range beta particles in 188Re can give this certainty that large tumors can be eradicated with high efficiency. On the other hand, 186Re with low energy, short range beta particles is adequate item to ruin small tumors with minimum side effects and high yield. Thus, each of these radioisotopes has features that can cover just part of the treatment individually. So we thought accompanying 186Re and 188Re must have the best outcome to treat tumors with various sizes. Irradiating natural rhenium with neutrons has this potential to produce parallel 186Re and 188R together. We are looking for investigating whether the natural rhenium irradiation, in addition the concurrent production of these radioisotopes, gives us the appropriate radioactivity values to produce compositional radiopharmaceuticals? In this research, the experimental and theoretical assessments of 186Re and 188R simultaneous production to reach compositional radiopharmaceutical by natural rhenium irradiation in the Tehran research reactor, as well as the type and amount of produced impurities have been investigated. The results showed that experimental data are in good agreement with theoretical calculations. The maximum relative error in data has been calculated 8%. The results showed that, in the simultaneous production 186Re and 188R via the natural rhenium irradiation method, the amounts of impurities are trivial compared to the main products, and the activities of main products are properly enough to produce compositional radiopharmaceuticals.

Basic and practical concepts of radiopharmaceutical purification methods.

The presence of radiochemical impurities in a radiopharmaceutical contributes to an unnecessary radiation burden for the patients or to an undesirable high radioactivity background, which reduces the imaging contrast or therapeutic efficacy. Therefore, if the radiolabeling process results in unsatisfactory radiochemical purity, a purification step is unavoidable. A successful purification process requires a profound knowledge about the radiopharmaceuticals of interest ranging from structural features to susceptibility to different conditions. Most radiopharmaceutical purification methods are based on solid-phase extraction (SPE), high-performance liquid chromatography (HPLC), size exclusion chromatography (SEC), ion-exchange chromatography (IEC), and liquid-liquid extraction (LLE). Here, we discuss the basic and applied concepts of these purifications methods as well as their advantages and limitations.

Evaluation of 72Se/72As generator and production of 72Se for supplying 72As as a potential PET imaging radionuclide.

Positron-emitting 72As is the PET imaging counterpart for beta-emitting 77As. Its parent, no carrier added (n.c.a.) 72Se, was produced for a 72Se/72As generator by irradiating an enriched 7°Ge metal-graphite target via the 70Ge(α, 2 n)72Se reaction. Target dissolution used a fast, environmentally friendly method with 93% radioactivity recovery. Chromatographic parameters of the 72Se/72As generator were evaluated, the eluted n.c.a. 72As was characterized with a phantom imaging study, and the previously reported trithiol and aryl-dithiol ligand systems were radiolabeled with the separated n.c.a. 72As in high yield.

A compact solvent extraction based 99Mo/99mTc generator for hospital radiopharmacy.

A compact and portable 99Mo-99 mTc generator based on solvent-extraction, mimic to the conventional 99Mo-99 mTc alumina column generator is much-needed commodity for use in hospital radiopharmacy setup. The present study includes the development of a portable, simple and low cost 99Mo/99 mTc-generator based on MEK solvent extraction technique to obtain a very high concentration of no-carrier added (nca) 99 mTc solution, where low specific activity 99Mo source is obtained through 98Mo(n, γ)99Mo reaction in a research reactor. The unit is intended for operation under the conditions of medical radiological laboratories. Technical trials showed that the mean time of preparation of sodium [99mTc] pertechnetate radiopharmaceutical did not exceed 15 min. The quality and yield of 99 mTc-pertechnetate is upto the mark for formulation of radiopharmaceuticals.

Production of Sc medical radioisotopes with proton and deuteron beams.

Proton and deuteron beams (15.3 and 6.8 MeV, respectively) extracted from the PETtrace medical cyclotron at the Radiopharmaceuticals Production and Research Centre in the University of Warsaw, Heavy Ion Laboratory, 28 MeV protons from the C30 cyclotron at the National Centre for Nuclear Research, Swierk, near Warsaw and 33 MeV protons from the ARRONAX accelerator, Nantes were used to produce and investigate the medically interesting Sc radioisotopes. Both natural and isotopically enriched CaCO3 and TiO2 targets were used (42Ca, 43Ca, 44Ca, 48Ca, 48Ti). The production efficiency and isotopic purity were determined and are reported here for the highest commercially available enrichments of the target material. The Thick Target Yield, Activities at the End of Bombardment (EOB) and the relative activities of produced impurities at EOB are reported for 43Sc, 44gSc, 44mSc and 47Sc produced with particle energies below 33 MeV.