Theoretical modeling of yields for proton-induced reactions on natural and enriched molybdenum targets

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.

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.

Overview and evaluation of different nuclear level density models for the 123I radionuclide production

The ¹²³I radionuclide (T_1/2 = 13.22 h, β+ = 100%) is one of the most potent gamma emitters for nuclear medicine. In this study, the cyclotron production of this radionuclide via different nuclear reactions namely, the ¹²¹Sb(α,2n), ¹²²Te(d,n), ¹²³Te(p,n), ¹²⁴Te(p,2n), ¹²⁴Xe(p,2n), ¹²⁷I(p,5n) and ¹²⁷I(d,6n) were investigated. The effect of the various phenomenological nuclear level density models such as Fermi gas model (FGM), Back-shifted Fermi gas model (BSFGM), Generalized superfluid model (GSM) and Enhanced generalized superfluid model (EGSM) moreover, the three microscopic level density models were evaluated for predicting of cross sections and production yield predictions. The SRIM code was used to obtain the target thickness. The ¹²³I excitation function of reactions were calculated by using of the TALYS-1.8, EMPIRE-3.2 nuclear codes and with data which taken from TENDL-2015 database, and finally the theoretical calculations were compared with reported experimental measurements in which taken from EXFOR database.

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.

Production of Y-86 and other radiometals for research purposes using a solution target system

INTRODUCTION

Diagnostic radiometals are typically obtained from cyclotrons by irradiating solid targets or from radioisotope generators. These methods have the advantage of high production yields, but require additional solid target handling infrastructure that is not readily available to many cyclotron facilities. Herein, we provide an overview of our results regarding the production of various positron-emitting radiometals using a liquid target system installed on a 13 MeV cyclotron at TRIUMF. Details about the production, purification and quality control of (89)Zr, (68)Ga and for the first time (86)Y are discussed.

METHODS

Aqueous solutions containing 1.35-1.65 g/mL of natural-abundance zinc nitrate, yttrium nitrate, and strontium nitrate were irradiated on a 13 MeV cyclotron using a standard liquid target. Different target body and foil materials were investigated for corrosion. Production yields were calculated using theoretical cross-sections from the EMPIRE code and compared with experimental results. The radioisotopes were extracted from irradiated target material using solid phase extraction methods adapted from previously reported methods, and used for radiolabelling experiments.

RESULTS

We demonstrated production quantities that are sufficient for chemical and biological studies for three separate radiometals, (89)Zr (Asat = 360 MBq/μA and yield = 3.17 MBq/μA), (86)Y (Asat = 31 MBq/μA and yield = 1.44 MBq/μA), and (68)Ga (Asat = 141 MBq/μA and yield = 64 MBq/μA) from one hour long irradiations on a typical medical cyclotron. (68)Ga yields were sufficient for potential clinical applications. In order to avoid corrosion of the target body and target foil, nitrate solutions were chosen as well as niobium as target-body material. An automatic loading system enabled up to three production runs per day. The separation efficiency ranged from 82 to 99%. Subsequently, (68)Ga and (86)Y were successfully used to radiolabel DOTA-based chelators while deferoxamine was used to coordinate (89)Zr.

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.

Graphical user interface for yield and dose estimations for cyclotron-produced technetium

The cyclotron-based (100)Mo(p,2n)(99m)Tc reaction has been proposed as an alternative method for solving the shortage of (99m)Tc. With this production method, however, even if highly enriched molybdenum is used, various radioactive and stable isotopes will be produced simultaneously with (99m)Tc. In order to optimize reaction parameters and estimate potential patient doses from radiotracers labeled with cyclotron produced (99m)Tc, the yields for all reaction products must be estimated. Such calculations, however, are extremely complex and time consuming. Therefore, the objective of this study was to design a graphical user interface (GUI) that would automate these calculations, facilitate analysis of the experimental data, and predict dosimetry. The resulting GUI, named Cyclotron production Yields and Dosimetry (CYD), is based on Matlab®. It has three parts providing (a) reaction yield calculations, (b) predictions of gamma emissions and (c) dosimetry estimations. The paper presents the outline of the GUI, lists the parameters that must be provided by the user, discusses the details of calculations and provides examples of the results. Our initial experience shows that the proposed GUI allows the user to very efficiently calculate the yields of reaction products and analyze gamma spectroscopy data. However, it is expected that the main advantage of this GUI will be at the later clinical stage when entering reaction parameters will allow the user to predict production yields and estimate radiation doses to patients for each particular cyclotron run.

The present and future of medical radionuclide production

Medical radionuclide production technology is well established. Both reactors and cyclotrons are utilized for production; the positron emitters, however, are produced exclusively using cyclotrons. A brief survey of the production methods of most commonly used diagnostic and therapeutic radionuclides is given. The emerging radionuclides are considered in more detail. They comprise novel positron emitters and therapeutic radionuclides emitting low-range electrons and α-particles. The possible alternative production routes of a few established radionuclides, like 68Ga and 99mTc, are discussed. The status of standardisation of production data of the commonly used as well as of some emerging radionuclides is briefly mentioned. Some notions on anticipated future trends in the production and application of radionuclides are considered.

A comparison of rat SPECT images obtained using (99m)Tc derived from 99Mo produced by an electron accelerator with that from a reactor

Recent shortages of molybdenum-99 ((99)Mo) have led to an examination of alternate production methods that could contribute to a more robust supply. An electron accelerator and the photoneutron reaction were used to produce (99)Mo from which technetium-99m ((99m)Tc) is extracted. SPECT images of rat anatomy obtained using the accelerator-produced (99m)Tc with those obtained using (99m)Tc from a commercial generator were compared. Disks of (100)Mo were irradiated with x-rays produced by a 35 MeV electron beam to generate about 1110 MBq (30 mCi) of (99)Mo per disk. After target dissolution, a NorthStar ARSII unit was used to separate the (99m)Tc, which was subsequently used to tag pharmaceuticals suitable for cardiac and bone imaging. SPECT images were acquired for three rats and compared to images for the same three rats obtained using (99m)Tc from a standard reactor (99)Mo generator. The efficiency of (99)Mo-(99m)Tc separation was typically greater than 90%. This study demonstrated the delivery of (99m)Tc from the end of beam to the end user of approximately 30 h. Images obtained using the heart and bone scanning agents using reactor and linac-produced (99m)Tc were comparable. High-power electron accelerators are an attractive option for producing (99)Mo on a national scale.

Evaluation ofMo99andTc99mProductions Based on a High-Performance Cyclotron

Following preliminary feasibility studies which started at Legnaro National Laboratories (LNL) in 2011, the Italian National Institute for Nuclear Physics (INFN) research activities are underway aiming at the alternative, accelerator-driven,Mo99/Tc99mproduction routes. One of the most promising approaches is to use100Mo-enriched (i.e., >99%) molybdenum metallic targets, bombarded with high-beam-current, high-energy proton cyclotrons. In order to get a comprehensive map of radionuclides expected, a detailed theoretical investigation has been carried out using the TALYS-TENDL 2012 excitation functions extended up to (p,6n), (p,p5n), and (p,2p4n) levels. A series of quality parameters have thus been calculated both at the end of beam (EOB) and at longer times. Results point out that accelerator-99Mo is of limited interest for a possible massive production because of the quite low specific activity with respect to reactor-99Mo. Accelerator-Tc99mquality parameters (i.e., radionuclidic purity (RNP), isotopic purity (IP), and specific activities) calculated are instead quite close to the generator-Tc. Calculations at 15, 20, and 25 MeV have thus been performed to assess the best operative irradiation condition forTc99mproduction while minimizing both the short-lived and long-lived Tc contaminant radionuclides. Although present in minimum quantities, Tc contaminants may indeed have an impact either on the pharmaceutical labeling procedures or on contributing to patient radiation dose during the diagnostic procedures.