Implementation of Multi-Curie Production of 99mTc by Conventional Medical Cyclotrons

Proceedings of international symposium of trends in radiopharmaceuticals 2023 (ISTR-2023)

The International Atomic Energy Agency (IAEA) held the 3rd International Symposium on Trends in Radiopharmaceuticals, (ISTR-2023) at IAEA Headquarters in Vienna, Austria, during the week of 16-21 April 2023. This procedural paper summarizes highlights from symposium presentations, posters, panel discussions and satellite meetings, and provides additional resources that may be useful to researchers working with diagnostic and therapeutic radiopharmaceuticals in the academic, government and industry setting amongst IAEA Member States and beyond. More than 550 participants in person from 88 Member States attended the ISTR-2023. Over 360 abstracts were presented from all over the world by a diverse group of global scientists working with radiopharmaceuticals. Given this group of international radiochemists is unique to ISTR (IAEA funding enabled many to attend), there was an invaluable wealth of knowledge on the global state of the radiopharmaceutical sciences present at the meeting. The intent of this Proceedings paper is to share this snapshot from our international colleagues with the broader radiopharmaceutical sciences community by highlighting presentations from the conference on the following topics: Isotope Production and Radiochemistry, Industrial Insights, Regional Trends, Training and Education, Women in the Radiopharmaceutical Sciences, and Future Perspectives and New Initiatives. The authors of this paper are employees of IAEA, members of the ISTR-2023 Organizing Committee and/or members of the EJNMMI Radiopharmacy and Chemistry Editorial Board who attended ISTR-2023. Overall, ISTR-2023 fostered the successful exchange of scientific ideas around every aspect of the radiopharmaceutical sciences. It was well attended by a diverse mix of radiopharmaceutical scientists from all over the world, and the oral and poster presentations provided a valuable update on the current state-of-the-art of the field amongst IAEA Member States. Presentations as well as networking amongst the attendees resulted in extensive knowledge transfer amongst the various stakeholders representing 88 IAEA Member States. This was considered particularly valuable for attendees from Member States where nuclear medicine and the radiopharmaceutical sciences are still relatively new. Since the goal is for the symposium series to be held every four years; the next one is anticipated to take place in 2027.

Preclinical PET Imaging and Toxicity Study of a 68Ga-Functionalized Polymeric Cardiac Blood Pool Agent

Cardiac blood pool imaging is currently performed almost exclusively with 99mTc-based compounds and SPECT/CT imaging. Using a generator-based PET radioisotope has a few advantages, including not needing nuclear reactors to produce it, obtaining better resolution in humans, and potentially reducing the radiation dose to the patient. When the shortlived radioisotope 68Ga is used, it can be applied repeatedly on the same day—for example, for the detection of bleeding. Our objective was to prepare and evaluate a long-circulating polymer functionalized with gallium for its biodistribution, toxicity, and dosimetric properties. A 500 kDa hyperbranched polyglycerol was conjugated to the chelator NOTA and radiolabeled rapidly at room temperature with 68Ga. It was then injected intravenously into a rat, and gated imaging allowed us to easily observe wall motion and cardiac contractility, confirming the suitability of this radiopharmaceutical for cardiac blood pool imaging. Internal radiation dose calculations showed that the radiation doses that patients would receive from the PET agent would be 2.5× lower than those from the 99mTc agent. A complete 14-day toxicology study in rats concluded that there were no gross pathology findings, changes in body or organ weights, or histopathological events. This radioactive-metal-functionalized polymer might be a suitable non-toxic agent to advance for clinical application.

Ultrafast and selective separation of 99mTc from molybdenum matrix using DBDGA deliberately tailored macrocyclic crown-ethers

Technetium-99m (^99mTc) is an important medical radionuclide. Due to the crisis in supply of molybdenum-99 (⁹⁹Mo), production of ^99mTc directly via the ¹⁰⁰Mo (p, 2 n) reaction by cyclotron was proposed. In this process, the most critical challenge is to rapidly and efficiently separate ^99mTc from high concentration of molybdenum. In this work, a novel ligand, bis(N,N-dibutyldiglycolamide)dibenzo-18-crown-6 (BisDBDGA-DB18C6) was successfully synthesized and used for extraction of TcO₄^- /ReO₄^- from molybdenum. The results demonstrated that BisDBDGA-DB18C6 expressed excellent selectivity for TcO₄^- with a high separation factor of 1.6 × 10⁵ against Mo, a fast extraction kinetic (within 45 s), and a high extraction capacity of 211 mmol ReO₄^- (⁹⁹TcO₄^-)/per mole of extractant. The extraction mechanism was proposed as a co-interaction of macrocyclic crown ether and N,N-dibutyldiglycolamide group through slope analysis, FT-IR, ESI-MS, ¹H NMR titration and theory calculations. Importantly, ⁹⁹Tc in the organic phase can be quantitatively (> 99%) and easily back-extracted using deionized water, which can be directly used for medical applications.

Application of therapeutic linear accelerators for the production of radioisotopes used in nuclear medicine

This review paper summarizes the possibilities of the use of therapeutic linear electron accelerators for the production of radioisotopes for nuclear medicine. This work is based on our published results and the thematically similar papers by other authors, directly related to five medical radioisotopes as 99Mo/99mTc, 198Au, 186Re, 188Re, 117mSn, produced using therapeutic linacs. Our unpublished data relating to the issues discussed have also been used here. In the experiments, two types of radiation were included in the analysis of the radioisotope production process, i.e. the therapeutic twenty-megavolt (20 MV) X-rays generated by Varian linacs and neutron radiation contaminating the therapeutic beam. Thus, the debated radioisotopes are produced in the photonuclear reactions and in the neutron ones. Linear therapeutic accelerators do not allow the production of radioisotopes with high specific activities, but the massive targets can be used instead. Thus, the amount of the produced radioisotopes may be increased. Apart from linear accelerators, more and more often, the production of radioisotopes is carried out in small medical cyclotrons. More such cyclotrons are developed, built, and sold commercially than for scientific research. The radioisotopes produced with the use of therapeutic linacs or cyclotrons can be successfully applied in various laboratory tests and in research.

Studies on electrochemical dissolution of sintered molybdenum discs as a potential method for targets dissolution in 99mTc production

Electrochemical dissolution of pressed into discs and sintered metallic molybdenum powder with the mass of 712 ± 10 mg (n = 15) in potassium hydroxide solution was studied in detail. The technique was considered to apply for dissolution of irradiated 100Mo target in the 99mTc production. The effect of various parameters, e.g., the concentration of the electrolyte solution, temperature, current density, and surface area of the platinum cathode, was investigated. The shortest time for total dissolution of molybdenum target was 70 min. This result was achieved using an electrolyte solution of 5 M KOH, temperature 55 °C and the current density of 365 mA/cm2.

A Universal Cassette-Based System for the Dissolution of Solid Targets

Cyclotron-based radionuclides production by using solid targets has become important in the last years due to the growing demand of radiometals, e.g., ⁶⁸Ga, ⁸⁹Zr, ^43/47Sc, and ^52/54Mn. This shifted the focus on solid target management, where the first fundamental step of the radiochemical processing is the target dissolution. Currently, this step is generally performed with commercial or home-made modules separated from the following purification/radiolabelling modules. The aim of this work is the realization of a flexible solid target dissolution system to be easily installed on commercial cassette-based synthesis modules. This would offer a complete target processing and radiopharmaceutical synthesis performable in a single module continuously. The presented solid target dissolution system concept relies on an open-bottomed vial positioned upon a target coin. In particular, the idea is to use the movement mechanism of a syringe pump to position the vial up and down on the target, and to exploit the heater/cooler reactor of the module as a target holder. All the steps can be remotely controlled and are incorporated in the cassette manifold together with the purification and radiolabelling steps. The performance of the device was tested by processing three different irradiated targets under different dissolution conditions.

Highly Efficient Micro-Scale Liquid-Liquid In-Flow Extraction of 99mTc from Molybdenum

The trend to achieve even more compact-sized systems is leading to the development of micro-scale reactors (lab-on-chip) in the field of radiochemical separation and radiopharmaceutical production. Technetium-99m extraction from both high and low specific activity molybdenum could be simply performed by MEK-driven solvent extraction if it were not for unpractical automation. The aim of this work is to develop a solvent extraction and separation process of technetium from molybdenum in a micro-scale in-flow chemistry regime with the aid of a capillary loop and a membrane-based separator, respectively. The developed system is able to extract and separate quantitatively and selectively (91.0 ± 1.8% decay corrected) the [^99mTc]TcO₄Na in about 20 min, by using a ZAIPUT separator device. In conclusion, we demonstrated for the first time in our knowledge the high efficiency of a MEK-based solvent extraction process of ^99mTc from a molybdenum-based liquid phased in an in-flow micro-scale regime.

A semi-automated module for the separation and purification of 99mTc from simulated molybdenum target

A semi-automated purification module for the cyclic separation of 99mTc was designed for production of [99mTc]TcO4– from γ irradiated 100Mo target. The separation process was carried out by using a 3-column purification system and the final product, [99mTc]TcO4–, was obtained in a total volume of 7 mL. To confirm proper separation achieved for 99mTc, a radio-labeling procedure using DTPA chelator was performed. The radiochemical purity was higher than 95%, which meets the strict radiopharmaceutical requirements. The yielded 99mTc can be separated with high efficiency from Mo in a quick and repeated way. Loss of 99mTc radioactivity during such a three-column separation process was not larger than 10%.

Recent Advances in Radiometals for Combined Imaging and Therapy in Cancer

Nuclear medicine is defined as the use of radionuclides for diagnostic and therapeutic applications. The imaging modalities positron emission tomography (PET) and single-photon emission computed tomography (SPECT) are based on γ-emissions of specific energies. The therapeutic technologies are based on β^- -particle-, α-particle-, and Auger electron emitters. In oncology, PET and SPECT are used to detect cancer lesions, to determine dosimetry, and to monitor therapy effectiveness. In contrast, radiotherapy is designed to irreparably damage tumor cells in order to eradicate or control the disease's progression. Radiometals are being explored for the development of diagnostic and therapeutic radiopharmaceuticals. Strategies that combine both modalities (diagnostic and therapeutic), referred to as theranostics, are promising candidates for clinical applications. This review provides an overview of the basic concepts behind therapeutic and diagnostic radiopharmaceuticals and their significance in contemporary oncology. Select radiometals that significantly impact current and upcoming cancer treatment strategies are grouped as clinically suitable theranostics pairs. The most important physical and chemical properties are discussed. Standard production methods and current radionuclide availability are provided to indicate whether a cost-efficient use in a clinical routine is feasible. Recent preclinical and clinical developments and outline perspectives for the radiometals are highlighted in each section.

Analysis of radioactive waste generated during the cyclotron production of 99mTc

Past and prospective shortages of medical radioisotopes have driven recent developments in the direct production of ^99mTc via the ¹⁰⁰Mo(p,2n)^99mTc reaction. The cyclotron-based production method has been shown to successfully produce ^99mTc, 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 ^99mTc 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 ^99mTc produced. Two-thirds of the total waste activity produced was attributed to ⁹⁹Mo (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 ^99mTc production was acceptably low for routine clinical production.

Innovative Target for Production of Technetium-99m by Biomedical Cyclotron

Technetium-99m (99mTc) is the most used radionuclide worldwide in nuclear medicine for diagnostic imaging procedures. 99mTc is typically extracted from portable generators containing 99Mo, which is produced normally in nuclear reactors as a fission product of highly enriched Uranium material. Due to unexpected outages or planned and unplanned reactor shutdown, significant 99mTc shortages appeared as a problem since 2008 The alternative cyclotron-based approach through the 100Mo(p,2n)99mTc reaction is considered one of the most promising routes for direct 99mTc production in order to mitigate potential 99Mo shortages. The design and manufacturing of appropriate cyclotron targets for the production of significant amounts of a radiopharmaceutical for medical use is a technological challenge. In this work, a novel solid target preparation method was developed, including sputter deposition of a dense, adherent, and non-oxidized Mo target material onto a complex backing plate. The latter included either chemically resistant sapphire or synthetic diamond brazed in vacuum conditions to copper. The target thermo-mechanical stability tests were performed under 15.6 MeV proton energy and different beam intensities, up to the maximum provided by the available GE Healthcare (Chicago, IL, USA) PET trace medical cyclotron. The targets resisted proton beam currents up to 60 µA (corresponding to a heat power density of about 1 kW/cm²) without damage or Mo deposited layer delamination. The chemical stability of the proposed backing materials was proven by gamma-spectroscopy analysis of the solution obtained after the standard dissolution procedure of irradiated targets in H₂O₂.

LARAMED: A Laboratory for Radioisotopes of Medical Interest

The widespread availability of novel radioactive isotopes showing nuclear characteristics suitable for diagnostic and therapeutic applications in nuclear medicine (NM) has experienced a great development in the last years, particularly as a result of key advancements of cyclotron-based radioisotope production technologies. At Legnaro National Laboratories of the National Institute of Nuclear Physics (LNL-INFN), Italy, a 70-MeV high current cyclotron has been recently installed. This cyclotron will be dedicated not only to pursuing fundamental nuclear physics studies, but also to research related to other scientific fields with an emphasis on medical applications. LARAMED project was established a few years ago at LNL-INFN as a new research line aimed at exploiting the scientific power of nuclear physics for developing innovative applications to medicine. The goal of this program is to elect LNL as a worldwide recognized hub for the development of production methods of novel medical radionuclides, still unavailable for the scientific and clinical community. Although the research facility is yet to become fully operative, the LARAMED team has already started working on the cyclotron production of conventional medical radionuclides, such as Tc-99m, and on emerging radionuclides of high potential medical interest, such as Cu-67, Sc-47, and Mn-52.

Antitumor Activity of Auger Electron Emitter 111In Delivered by Modular Nanotransporter for Treatment of Bladder Cancer With EGFR Overexpression

Gamma-ray emitting ¹¹¹In, which is extensively used for imaging, is also a source of short-range Auger electrons (AE). While exhibiting negligible effect outside cells, these AE become highly toxic near DNA within the cell nucleus. Therefore, these radionuclides can be used as a therapeutic anticancer agent if delivered precisely into the nuclei of tumor target cells. Modular nanotransporters (MNTs) designed to provide receptor-targeted delivery of short-range therapeutic cargoes into the nuclei of target cells are perspective candidates for specific intracellular delivery of AE emitters. The objective of this study was to evaluate the in vitro and in vivo efficacy of ¹¹¹In attached MNTs to kill human bladder cancer cells overexpressing epidermal growth factor receptor (EGFR). The cytotoxicity of ¹¹¹In delivered by the EGFR-targeted MNT (¹¹¹In-MNT) was greatly enhanced on EJ-, HT-1376-, and 5637-expressing EGFR bladder cancer cell lines compared with ¹¹¹In non-targeted control. In vivo microSPECT/CT imaging and antitumor efficacy studies revealed prolonged intratumoral retention of ¹¹¹In-MNT with t½ = 4.1 ± 0.5 days as well as significant dose-dependent tumor growth delay (up to 90% growth inhibition) after local infusion of ¹¹¹In-MNT in EJ xenograft-bearing mice.

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.

Radioactive Transition Metals for Imaging and Therapy

Nuclear medicine is composed of two complementary areas, imaging and therapy. Positron emission tomography (PET) and single-photon imaging, including single-photon emission computed tomography (SPECT), comprise the imaging component of nuclear medicine. These areas are distinct in that they exploit different nuclear decay processes and also different imaging technologies. In PET, images are created from the 511 keV photons produced when the positron emitted by a radionuclide encounters an electron and is annihilated. In contrast, in single-photon imaging, images are created from the γ rays (and occasionally X-rays) directly emitted by the nucleus. Therapeutic nuclear medicine uses particulate radiation such as Auger or conversion electrons or β^- or α particles. All three of these technologies are linked by the requirement that the radionuclide must be attached to a suitable vector that can deliver it to its target. It is imperative that the radionuclide remain attached to the vector before it is delivered to its target as well as after it reaches its target or else the resulting image (or therapeutic outcome) will not reflect the biological process of interest. Radiochemistry is at the core of this process, and radiometals offer radiopharmaceutical chemists a tremendous range of options with which to accomplish these goals. They also offer a wide range of options in terms of radionuclide half-lives and emission properties, providing the ability to carefully match the decay properties with the desired outcome. This Review provides an overview of some of the ways this can be accomplished as well as several historical examples of some of the limitations of earlier metalloradiopharmaceuticals and the ways that new technologies, primarily related to radionuclide production, have provided solutions to these problems.

In-house cyclotron production of high-purity Tc-99m and Tc-99m radiopharmaceuticals

In the last years, the technology for producing the important medical radionuclide technetium-99m by cyclotrons has become sufficiently mature to justify its introduction as an alternative source of the starting precursor [^99mTc][TcO₄]^- ubiquitously employed for the production of ^99mTc-radiopharmaceuticals in hospitals. These technologies make use almost exclusively of the nuclear reaction ¹⁰⁰Mo(p,2n)^99mTc that allows direct production of Tc-99m. In this study, it is conjectured that this alternative production route will not replace the current supply chain based on the distribution of ⁹⁹Mo/^99mTc generators, but could become a convenient emergency source of Tc-99m only for in-house hospitals equipped with a conventional, low-energy, medical cyclotron. On this ground, an outline of the essential steps that should be implemented for setting up a hospital radiopharmacy aimed at the occasional production of Tc-99m by a small cyclotron is discussed. These include (1) target production, (2) irradiation conditions, (3) separation/purification procedures, (4) terminal sterilization, (5) quality control, and (6) Mo-100 recovery. To address these issues, a comprehensive technology for cyclotron-production of Tc-99m, developed at the Legnaro National Laboratories of the Italian National Institute of Nuclear Physics (LNL-INFN), will be used as a reference example.

Production of novel diagnostic radionuclides in small medical cyclotrons

The global network of cyclotrons has expanded rapidly over the last decade. The bulk of its industrial potential is composed of small medical cyclotrons with a proton energy below 20 MeV for radionuclides production. This review focuses on the recent developments of novel medical radionuclides produced by cyclotrons in the energy range of 3 MeV to 20 MeV. The production of the following medical radionuclides will be described based on available literature sources: Tc-99 m, I-123, I-124, Zr-89, Cu-64, Ga-67, Ga-68, In-111, Y-86 and Sc-44. Remarkable developments in the production process have been observed in only some cases. More research is needed to make novel radionuclide cyclotron production available for the medical industry.

Validating production of PET radionuclides in solid and liquid targets: Comparing Geant4 predictions with FLUKA and measurements

The Monte Carlo toolkit Geant4 is used to simulate the production of a number of positron emitting radionuclides: ¹³N, ¹⁸F, ⁴⁴Sc, ⁵²Mn, ⁵⁵Co ⁶¹Cu, ⁶⁸Ga, ⁸⁶Y, ⁸⁹Zr and ⁹⁴Tc, which have been produced using a 13MeV medical cyclotron. The results are compared to previous simulations with the Monte Carlo code FLUKA and experimental measurements. The comparison shows variable degrees of agreement for different isotopes. The mean absolute deviation of Monte Carlo results from experiments was 1.4±1.6 for FLUKA and 0.7±0.5 for Geant4 using TENDL cross sections with QGSP-BIC-AllHP physics. Both agree well within the large error, which is due to the uncertainties present in both experimentally determined and theoretical reaction cross sections. Overall, Geant4 has been confirmed as a tool to simulate radionuclide production at low proton energy.

Cyclotron production of 99mTc: Comparison of known separation technologies for isolation of 99mTc from molybdenum targets

Intensive efforts were undertaken during the last few decades for the separation of cyclotron-produced ^99mTc from ⁹⁹Mo 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 ¹⁰⁰Mo(p,2n)^99mTc appears to be the most worthwhile approach. New ways of producing of ^99mTc require efficient separation methods. Several strategies for separation of ^99mTc from ⁹⁹Mo have been already developed. The advantages, disadvantages and technical challenges toward application potential of investigated methods to separate ^99mTc from irradiated ¹⁰⁰Mo target are discussed. These methods include column chromatography, solvent extraction, chemical precipitation and thermochromatography.

Impurities in Tc-99m radiopharmaceutical solution obtained from Mo-100 in cyclotron

The gamma emitting impurities in ^99mTc solution obtained from enriched molybdenum ¹⁰⁰Mo metallic target after its irradiation in a cyclotron were measured using a high-purity germanium (HPGe) detector. The radioactivity range of tested samples of ^99mTc was rather low, in the range from 0.34 to 2.39 MBq, thus creating a challenge to investigate the standard measurement HPGe system for impurity detection and quantification. In the process of ^99mTc separation from irradiated target the AnaLig® Tc-02 resin, Dionex H⁺ and Alumina A columns were used. Fractions of eluates from various steps of separation process were taken and measured for radionuclidic purity. The overall measurement sensitivity of gamma emitters in terms of minimum detectable activity (MDA) was found at the level of 14-70Bq with emission lines in range of 36 - 1836keV resulting in impurity content range of 6.7 × 10^-4 to 3.4 × 10^-3 % for ⁹³Tc, ^93mTc, ⁹⁴Tc, ^94mTc, ⁹⁵Tc, ^95mTc, ⁹⁶Tc ⁹⁶Nb, ⁹⁷Nb, ⁹⁹Mo contaminants and 9.4 × 10^-3 % for ^97mTc. The usefulness of the chosen measurement conditions and the method applied to testing the potential contaminators was proved by reaching satisfactory results of MDAs less than the criteria of impurity concentration of all nuclides specified in the European Pharmacopoeia.

Recent achievements in Tc-99m radiopharmaceutical direct production by medical cyclotrons

^99mTc is the most commonly used radionuclide in the field of diagnostic imaging, a noninvasive method intended to diagnose a disease, assess the disease state and monitor the effects of treatments. Annually, the use of ^99mTc, covers about 85% of nuclear medicine applications. This isotope releases gamma rays at about the same wavelength as conventional X-ray diagnostic equipment, and owing to its short half-life (t_½ = 6 h) is ideal for diagnostic nuclear imaging. A patient can be injected with a small amount of ^99mTc and within 24 h almost 94% of the injected radionuclide would have decayed and left the body, limiting the patient's radiation exposure. ^99mTc is usually supplied to hospitals through a ⁹⁹Mo/^99mTc radionuclide generator system where it is produced from the β decay of the parent nuclide ⁹⁹Mo (t_½ = 66 h), which is produced in nuclear reactors via neutron fission. Recently, the interruption of the global supply chain of reactor-produced ⁹⁹Mo, has forced the scientific community to investigate alternative production routes for ^99mTc. One solution was to consider cyclotron-based methods as potential replacement of reactor-based technology and the nuclear reaction ¹⁰⁰Mo(p,2n)^99mTc emerged as the most worthwhile approach. This review reports some achievements about ^99mTc produced by medical cyclotrons. In particular, the available technologies for target design, the most efficient extraction and separation procedure developed for the purification of ^99mTc from the irradiated targets, the preparation of high purity ^99mTc radiopharmaceuticals and the first clinical studies carried out with cyclotron produced ^99mTc are described.

Manufacturing and characterization of molybdenum pellets used as targets for 99mTc production in cyclotron

The method of ¹⁰⁰Mo metallic target preparation for production of ^99mTc by proton irradiation in ¹⁰⁰Mo(p,2n)^99mTc reaction was demonstrated. For this purpose, pressing of molybdenum powder into pellets and their subsequent sintering in reductive atmosphere were applied. The influence of parameters such as molybdenum mass and time of both pressing and sintering on the ¹⁰⁰Mo target durability was investigated. Under the optimized conditions, ¹⁰⁰Mo metallic pellet targets with density of 9.95±0.06g/cm³ were obtained. Morphology and structure of pressed pellets before and after sintering were studied by using standard optical microscope and Scanning Electron Microscope (SEM). Nanoindentation technique was used to investigate the mechanical properties such as nanohardness and Young modulus. Prepared ¹⁰⁰Mo pellets were successfully irradiated with protons and ^99mTc was efficiently isolated.

Imaging study of using radiopharmaceuticals labeled with cyclotron-produced 99mTc

Cyclotron-produced ^99mTc (CPTc) has been recognized as an attractive and practical substitution of reactor/generator based ^99mTc. However, the small amount of ^92-98Mo in the irradiation of enriched ¹⁰⁰Mo could lead to the production of other radioactive technetium isotopes (Tc-impurities) which cannot be chemically separated. Thus, these impurities could contribute to patient dose and affect image quality. The potential radiation dose caused by these Tc-impurities produced using different targets, irradiation conditions, and corresponding to different injection times have been investigated, leading us to create dose-based limits of these parameters for producing clinically acceptable CPTc. However, image quality has been not considered. The aim of the present work is to provide a comprehensive and quantitative analysis of image quality for CPTc. The impact of Tc-impurities in CPTc on image resolution, background noise, and contrast is investigated by performing both Monte-Carlo simulations and phantom experiments. Various targets, irradiation, and acquisition conditions are employed for investigating the image-based limits of CPTc production parameters. Additionally, the relationship between patient dose and image quality of CPTc samples is studied. Only those samples which meet both dose- and image-based limits should be accepted in future clinical studies.

Application of AnaLig resin for (99m)Tc separation from (100)Mo target irradiated in cyclotron

The purpose of this study was the development of procedure for molybdenum metallic target processing after its irradiation in a cyclotron. As a first step the dissolution of molybdenum in various physical forms was investigated. The concentrations of NaOH and (NH4)2CO3 allowing the highest sorption of Tc on AnaLig Tc-02 resin had been found. Based on these results the sintered irradiated Mo pellet was processed. The radionuclidic and radiochemical purities of separated Tc product were evaluated.

Diversification of 99Mo/99mTc separation: non–fission reactor production of 99Mo as a strategy for enhancing 99mTc availability

This paper discusses the benefits of obtaining (99m)Tc from non-fission reactor-produced low-specific-activity (99)Mo. This scenario is based on establishing a diversified chain of facilities for the distribution of (99m)Tc separated from reactor-produced (99)Mo by (n,γ) activation of natural or enriched Mo. Such facilities have expected lower investments than required for the proposed chain of cyclotrons for the production of (99m)Tc. Facilities can receive and process reactor-irradiated Mo targets then used for extraction of (99m)Tc over a period of 2 wk, with 3 extractions on the same day. Estimates suggest that a center receiving 1.85 TBq (50 Ci) of (99)Mo once every 4 d can provide 1.48-3.33 TBq (40-90 Ci) of (99m)Tc daily. This model can use research reactors operating in the United States to supply current (99)Mo needs by applying natural (nat)Mo targets. (99)Mo production capacity can be enhanced by using (98)Mo-enriched targets. The proposed model reduces the loss of (99)Mo by decay and avoids proliferation as well as waste management issues associated with fission-produced (99)Mo.

A generator-produced gallium-68 radiopharmaceutical for PET imaging of myocardial perfusion

Lipophilic cationic technetium-99m-complexes are widely used for myocardial perfusion imaging (MPI). However, inherent uncertainties in the supply chain of molybdenum-99, the parent isotope required for manufacturing ⁹⁹Mo/^99mTc generators, intensifies the need for discovery of novel MPI agents incorporating alternative radionuclides. Recently, germanium/gallium (Ge/Ga) generators capable of producing high quality ⁶⁸Ga, an isotope with excellent emission characteristics for clinical PET imaging, have emerged. Herein, we report a novel ⁶⁸Ga-complex identified through mechanism-based cell screening that holds promise as a generator-produced radiopharmaceutical for PET MPI.

Cross-linked polyethylene glycol beads to separate 99mTc-pertechnetate from low-specific-activity molybdenum

We report a kit-based approach for the purification of sodium pertechnetate ((99m)TcO4 (-)) from solutions with high MoO4 (2-) content.

METHODS

Cross-linked polyethylene glycol resins (ChemMatrix) were used to separate (99m)Tc and molybdenum in 4N NaOH. The resins were loaded at various flow rates and eluted with water to release (99m)Tc. The (99m)Tc solution was passed through a cation exchange resin and an alumina cartridge, followed by saline elution. This process was tested with cyclotron-produced (99m)Tc using an automated system and disposable kits.

RESULTS

Optimal results were obtained by loading 500 mg of resin at flow rates of up to 3.1 mL/min, with quantitative extraction of (99m)Tc from the molybdate solution and complete release of (99m)Tc after elution with water. The automated system was highly efficient at isolating Na(99m)TcO4 within minutes, with a recovery rate of 92.7% ± 1.1% (mean ± SD) using cyclotron-produced (99m)Tc.

CONCLUSION

ChemMatrix resins were highly effective at separating (99m)TcO4 (-) from molybdate solutions.