Study of activation cross sections of double neutron capture reaction on 193Ir for the reactor production route of radiotherapeutic 195mPt

The radiotherapeutic 195mPt is among the most effective Auger electron emitters of the currently studied radionuclides that have a potential theranostic application in nuclear medicine. Production of 195mPt through double neuron capture of enriched 193Ir followed by β--decay to the radioisotope of interest carried out at the research reactor IBR-2 is described. Because of the high radiation background, radiochemical purification procedure of 195mPt from bulk of iridium was needed to be developed and is detailed here as well. For the first time, cross section and resonance integral for the reaction 194Ir(n,γ)195mIr were determined. Resonance neutrons contribution was established to exceed that of thermal neutrons, and resonance integral for the reaction 194Ir(n,γ)195mIr is calculated to be 2900 b. Specific activity of 195mPt was estimated to reach a value of 38.7 GBq/(g Pt) at IBR-2 by the end of bombardment (EOB).

Chromatographic separation of silver-111 from neutron-irradiated palladium target: toward direct labeling of radiotracers

BACKGROUND

Silver-111 is a promising β--emitting radioisotope with ideal characteristics for targeted radionuclide therapy and associated single photon emission tomography imaging. Its decay properties closely resemble the clinically established lutetium-177, making it an attractive candidate for therapeutic applications. In addition, the clinical value of silver-111 is further enhanced by the existence of the positron-emitting counterpart silver-103, thus imparting a truly theranostic potential to this element. A so-fitting matching pair could potentially overcome the current limitations associated with the forced use of chemically different isotopes as imaging surrogates of lutetium-177, leading to more accurate and efficient diagnosis and treatment. However, the use of silver-111-based radiopharmaceuticals in vivo has faced obstacles due to the challenges related to its production and radiochemical separation from the target material. To address these issues, this study aims to implement a chromatographic separation methodology for the purification of reactor-produced silver-111. The ultimate goal is to achieve a ready-to-use formulation for the direct radiolabeling of tumour-seeking biomolecules.

RESULTS

A two-step sequence chromatographic process was validated for cold Ag-Pd separation and then translated to the radioactive counterpart. Silver-111 was produced via the 110Pd(n,γ)111Pd nuclear reaction on a natural palladium target and the subsequent β--decay of palladium-111. Silver-111 was chemically separated from the metallic target via the implemented chromatographic process by using commercially available LN and TK200 resins. The effectiveness of the separations was assessed by inductively coupled plasma optical emission spectroscopy and γ-spectrometry, respectively, and the Ag+ retrieval was afforded in pure water. Recovery of silver-111 was > 90% with a radionuclidic purity > 99% and a separation factor of around 4.21·10-4.

CONCLUSIONS

The developed separation method was suitable to obtain silver-111 with high molar activity in a ready-to-use water-based formulation that can be directly employed for the labeling of radiotracers. By successfully establishing a robust and efficient production and purification method for silver-111, this research paves the way for its wider application in targeted radionuclide therapy and precision imaging.

Cutting edge rare earth radiometals: prospects for cancer theranostics

Background

With recent advances in novel approaches to cancer therapy and imaging, the application of theranostic techniques in personalised medicine has emerged as a very promising avenue of research inquiry in recent years. Interest has been directed towards the theranostic potential of Rare Earth radiometals due to their closely related chemical properties which allow for their facile and interchangeable incorporation into identical bifunctional chelators or targeting biomolecules for use in a diverse range of cancer imaging and therapeutic applications without additional modification, i.e. a “one-size-fits-all” approach. This review will focus on recent progress and innovations in the area of Rare Earth radionuclides for theranostic applications by providing a detailed snapshot of their current state of production by means of nuclear reactions, subsequent promising theranostic capabilities in the clinic, as well as a discussion of factors that have impacted upon their progress through the theranostic drug development pipeline.

Main body

In light of this interest, a great deal of research has also been focussed towards certain under-utilised Rare Earth radionuclides with diverse and favourable decay characteristics which span the broad spectrum of most cancer imaging and therapeutic applications, with potential nuclides suitable for α-therapy (¹⁴⁹Tb), β⁻-therapy (⁴⁷Sc, ¹⁶¹Tb, ¹⁶⁶Ho, ¹⁵³Sm, ¹⁶⁹Er, ¹⁴⁹Pm, ¹⁴³Pr, ¹⁷⁰Tm), Auger electron (AE) therapy (¹⁶¹Tb, ¹³⁵La, ¹⁶⁵Er), positron emission tomography (⁴³Sc, ⁴⁴Sc, ¹⁴⁹Tb, ¹⁵²Tb, ¹³²La, ¹³³La), and single photon emission computed tomography (⁴⁷Sc, ¹⁵⁵Tb, ¹⁵²Tb, ¹⁶¹Tb, ¹⁶⁶Ho, ¹⁵³Sm, ¹⁴⁹Pm, ¹⁷⁰Tm). For a number of the aforementioned radionuclides, their progression from ‘bench to bedside’ has been hamstrung by lack of availability due to production and purification methods requiring further optimisation.

Conclusions

In order to exploit the potential of these radionuclides, reliable and economical production and purification methods that provide the desired radionuclides in high yield and purity are required. With more reactors around the world being decommissioned in future, solutions to radionuclide production issues will likely be found in a greater focus on linear accelerator and cyclotron infrastructure and production methods, as well as mass separation methods. Recent progress towards the optimisation of these and other radionuclide production and purification methods has increased the feasibility of utilising Rare Earth radiometals in both preclinical and clinical settings, thereby placing them at the forefront of radiometals research for cancer theranostics.

Current status on cyclotron facilities and related infrastructure supporting PET applications in Latin America and the Caribbean

This review presents the results of a survey conducted by the International Atomic Energy Agency on cyclotrons and related infrastructure used for radionuclide and radiopharmaceutical production which are supporting PET imaging applications in Latin America and the Caribbean region.

Target manufacturing by Spark Plasma Sintering for efficient 89Zr production

Zirconium-89 (⁸⁹Zr) is an emerging radionuclide for positron emission tomography (PET), with nuclear properties suitable for imaging slow biological processes in cellular targets. The ⁸⁹Y(p,n)⁸⁹Zr nuclear reaction is commonly exploited as the main production route with medical cyclotrons accelerating low-energy (< 20 MeV) and low-current (< 100 μA) proton beams. Usually, natural yttrium solid targets manufactured by different methods, including yttrium electrodeposition, yttrium sputtering, compressed yttrium powders, and foils, were employed. In this study, the Spark Plasma Sintering (SPS) technique has been investigated, for the first time, to manufacture yttrium solid targets for an efficient ⁸⁹Zr radionuclide yield. The natural yttrium disc was bonded to a niobium backing plate using a commercial SPS apparatus and a prototype machine assembled at the University of Pavia. The resulting targets were irradiated in a TR19 cyclotron with a 12 MeV proton beam at 50 μA. A dedicated dissolution module, obtained from a commercial system, was used to develop an automated process for the purification and recovery of the produced ⁸⁹Zr radionuclide. The production yield and recovery efficiency were measured and compared to ⁸⁹Zr produced by irradiating standard yttrium foils. SPS manufactured targets withstand an average heat power density of approximately 650 W∙cm^-2 for continuous irradiation up to 5 h without visible damage. A saturation yield of 14.12 ± 0.38 MBq/μAh was measured. The results showed that the obtained ⁸⁹Zr production yield and quality were comparable to similar data obtained using standard yttrium foil targets. In conclusion, the present work demonstrates that the SPS technique might be a suitable technical manufacturing solution aimed at high-yield ⁸⁹Zr radioisotope production.

Cyclotron production and radiochemical purification of terbium-155 for SPECT imaging

Background

Terbium-155 [T_1/2 = 5.32 d, Eγ = 87 keV (32%) 105 keV (25%)] is an interesting radionuclide suitable for single photon emission computed tomography (SPECT) imaging with potential application in the diagnosis of oncological disease. It shows similar decay characteristics to the clinically established indium-111 and would be a useful substitute for the diagnosis and prospective dosimetry with biomolecules that are afterwards labeled with therapeutic radiolanthanides and pseudo-radiolanthanides, such as lutetium-177 and yttrium-90. Moreover, terbium-155 could form part of the perfect “matched pair” with the therapeutic radionuclide terbium-161, making the concept of true radiotheragnostics a reality. The aim of this study was the investigation of the production of terbium-155 via the ¹⁵⁵Gd(p,n)¹⁵⁵Tb and ¹⁵⁶Gd(p,2n)¹⁵⁵Tb nuclear reactions and its subsequent purification, in order to obtain a final product in quantity and quality sufficient for preclinical application. The ¹⁵⁶Gd(p,2n)¹⁵⁵Tb nuclear reaction was performed with 72 MeV protons (degraded to ~ 23 MeV), while the ¹⁵⁵Gd(p,n)¹⁵⁵Tb reaction was degraded further to ~ 10 MeV, as well as performed at an 18 MeV medical cyclotron, to demonstrate its feasibility of production.

Result

The ¹⁵⁶Gd(p,2n)¹⁵⁵Tb nuclear reaction demonstrated higher production yields of up to 1.7 GBq, however, lower radionuclidic purity when compared to the final product (~ 200 MBq) of the ¹⁵⁵Gd(p,n)¹⁵⁵Tb nuclear reaction. In particular, other radioisotopes of terbium were produced as side products. The radiochemical purification of terbium-155 from the target material was developed to provide up to 1.0 GBq product in a small volume (~ 1 mL 0.05 M HCl), suitable for radiolabeling purposes. The high chemical purity of terbium-155 was proven by radiolabeling experiments at molar activities up to 100 MBq/nmol. SPECT/CT experiments were performed in tumor-bearing mice using [¹⁵⁵Tb]Tb-DOTATOC.

Conclusion

This study demonstrated two possible production routes for high activities of terbium-155 using a cyclotron, indicating that the radionuclide is more accessible than the exclusive mass-separated method previously demonstrated. The developed radiochemical purification of terbium-155 from the target material yielded [¹⁵⁵Tb]TbCl₃ in high chemical purity. As a result, initial cell uptake investigations, as well as SPECT/CT in vivo studies with [¹⁵⁵Tb]Tb-DOTATOC, were successfully performed, indicating that the chemical separation produced a product with suitable quality for preclinical studies.

Supplementary Information

The online version contains supplementary material available at 10.1186/s41181-021-00153-w.

Production, purification and availability of 211At: Near term steps towards global access

The promising characteristics of the 7.2-h radiohalogen ²¹¹At have long been recognized; including having chemical properties suitable for labeling targeting vectors ranging from small organic molecules to proteins, and the emission of only one α-particle per decay, providing greater control over off-target effects. Unfortunately, the impact of ²¹¹At within the targeted α-particle therapy domain has been constrained by its limited availability. Paradoxically, the most commonly used production method - via the ²⁰⁹Bi(α,2n)²¹¹At reaction - utilizes a widely available natural material (bismuth) as the target and straightforward cyclotron irradiation methodology. On the other hand, the most significant impediment to widespread ²¹¹At availability is the need for an accelerator capable of generating ≥28 MeV α-particles with sufficient beam intensities to make clinically relevant levels of ²¹¹At. In this review, current methodologies for the production and purification of ²¹¹At - both by the direct production route noted above and via a ²¹¹Rn generator system - will be discussed. The capabilities of cyclotrons that currently produce ²¹¹At will be summarized and the characteristics of other accelerators that could be utilized for this purpose will be described. Finally, the logistics of networks, both academic and commercial, for facilitating ²¹¹At distribution to locations remote from production sites will be addressed.

Characterization of actinide resin for separation of 51,52gMn from bulk target material

We report an extraction chromatography-based method via Actinide Resin for the isolation of radio-manganese from both natural chromium and isotopically enriched iron targets for cyclotron production of ^52gMn and ⁵¹Mn. For the separation of ^52gMn from n^atCr, a decay-corrected radiochemical yield of 83.7 ± 8.4% was achieved. For ⁵¹Mn from ⁵⁴Fe, a decay-corrected radiochemical yield of 78 ± 11% was achieved. This automatable method efficiently isolates both radionuclides from accelerator target material.

Expanding PET-applications in life sciences with positron-emitters beyond fluorine-18

Positron-emission-tomography (PET) has become an indispensable diagnostic tool in modern nuclear medicine. Its outstanding molecular imaging features allow repetitive studies on one individual and with high sensitivity, though no interference. Rather few positron-emitters with near favourable physical properties, i.e. carbon-11 and fluorine-18, furnished most studies in the beginning, preferably if covalently bound as isotopic label of small molecules. With the advancement of PET-devices the scope of in vivo research in life sciences and especially that of medical applications expanded, and other than "standard" PET-nuclides received increasing significance, like the radiometals copper-64 and gallium-68. Especially during the last decades, positron-emitters of other chemical elements have gotten into the focus of interest, concomitant with the technical advancements in imaging and radionuclide production. With known nuclear imaging properties and main production methods of emerging positron-emitters their usefulness for medical application is promising and even proven for several ones already. Unfortunate decay properties could be corrected for, and β⁺-emitters, especially with a longer half-life, provided new possibilities for application where slower processes are of importance. Further on, (bio)chemical features of positron-emitters of other elements, among there many metals, not only expanded the field of classical clinical investigations, but also opened up new fields of application. Appropriately labelled peptides, proteins and nanoparticles lend itself as newer probes for PET-imaging, e.g. in theragnostic or PET/MR hybrid imaging. Furthermore, the potential of non-destructive in-vivo imaging with positron-emission-tomography directs the view on further areas of life sciences. Thus, exploiting the excellent methodology for basic research on molecular biochemical functions and processes is increasingly encouraged as well in areas outside of health, such as plant and environmental sciences.

Evaluation of polydentate picolinic acid chelating ligands and an α-melanocyte-stimulating hormone derivative for targeted alpha therapy using ISOL-produced 225Ac

BACKGROUND

Actinium-225 (²²⁵Ac, t_1/2 = 9.9 d) is a promising candidate radionuclide for use in targeted alpha therapy (TAT), though the currently limited global supply has hindered the development of a suitable Ac-chelating ligand and ²²⁵Ac-radiopharmaceuticals towards the clinic. We at TRIUMF have leveraged our Isotope Separation On-Line (ISOL) facility to produce ²²⁵Ac and use the resulting radioactivity to screen a number of potential ²²⁵Ac-radiopharmaceutical compounds.

RESULTS

MBq quantities of ²²⁵Ac and parent radium-225 (²²⁵Ra, t_1/2 = 14.8 d) were produced and separated using solid phase extraction DGA resin, resulting in a radiochemically pure ²²⁵Ac product in > 98% yield and in an amenable form for radiolabeling of ligands and bioconjugates. Of the many polydentate picolinic acid ("pa") containing ligands evaluated (H₄octapa [N₄O₄], H₄CHXoctapa [N₄O₄], p-NO₂-Bn-H₄neunpa [N₅O₄], and H₆phospa [N₄O₄]), all out-performed the current gold standard, DOTA for ²²⁵Ac radiolabeling ability at ambient temperature. Moreover, a melanocortin 1 receptor-targeting peptide conjugate, DOTA-modified cyclized α-melanocyte-stimulating hormone (DOTA-CycMSH), was radiolabeled with ²²⁵Ac and proof-of-principle biodistribution studies using B16F10 tumour-bearing mice were conducted. At 2 h post-injection, tumour-to-blood ratios of 20.4 ± 3.4 and 4.8 ± 2.4 were obtained for the non-blocking (molar activity [M.A.] > 200 kBq/nmol) and blocking (M.A. = 1.6 kBq/nmol) experiment, respectively.

CONCLUSION

TRIUMF's ISOL facility is able to provide ²²⁵Ac suitable for preclinical screening of radiopharmaceutical compounds; [²²⁵Ac(octapa)]^-, [²²⁵Ac(CHXoctapa)]^-, and [²²⁵Ac(DOTA-CycMSH)] may be good candidates for further targeted alpha therapy studies.

Production of Ga-68 with a General Electric PETtrace cyclotron by liquid target

PURPOSE

In recent years the use of ⁶⁸Ga (t_1/2 = 67.84 min, β⁺: 88.88%) for the labelling of different PET radiopharmaceuticals has significantly increased. This work aims to evaluate the feasibility of the production of ⁶⁸Ga via the ⁶⁸Zn(p,n)⁶⁸Ga reaction by proton irradiation of an enriched zinc solution, using a biomedical cyclotron, in order to satisfy its increasing demand.

METHODS

Irradiations of 1.7 Msolution of ⁶⁸Zn(NO₃)₂ in 0.2 N HNO₃ were conducted with a GE PETtrace cyclotron using a slightly modified version of the liquid target used for the production of fluorine-18. The proton beam energy was degraded to 12 MeV, in order to minimize the production of ⁶⁷Ga through the⁶⁸Zn(p,2n)⁶⁷Ga reaction. The product's activity was measured using a calibrated activity meter and a High Purity Germanium gamma-ray detector.

RESULTS

The saturation yield of⁶⁸Ga amounts to (330 ± 20) MBq/µA, corresponding to a produced activity of⁶⁸Ga at the EOB of (4.3 ± 0.3) GBq in a typical production run at 46 µA for 32 min. The radionuclidic purity of the⁶⁸Ga in the final product, after the separation, is within the limits of the European Pharmacopoeia (>99.9%) up to 3 h after the EOB. Radiochemical separation up to a yield not lower than 75% was obtained using an automated purification module. The enriched material recovery efficiency resulted higher than 80-90%.

CONCLUSIONS

In summary, this approach provides clinically relevant amounts of⁶⁸Ga by cyclotron irradiation of a liquid target, as a competitive alternative to the current production through the⁶⁸Ge/⁶⁸Ga generators.

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.

Production and separation of 43Sc for radiopharmaceutical purposes

Background

The favorable decay properties of ⁴³Sc and ⁴⁴Sc for PET make them promising candidates for future applications in nuclear medicine. An advantage ⁴³Sc (T_1/2 = 3.89 h, Eβ⁺_av = 476 keV [88%]) exhibits over ⁴⁴Sc, however, is the absence of co-emitted high energy γ-rays. While the production and application of ⁴⁴Sc has been comprehensively discussed, research concerning ⁴³Sc is still in its infancy. This study aimed at developing two different production routes for ⁴³Sc, based on proton irradiation of enriched ⁴⁶Ti and ⁴³Ca target material.

Results

⁴³Sc was produced via the ⁴⁶Ti(p,α)⁴³Sc and ⁴³Ca(p,n)⁴³Sc nuclear reactions, yielding activities of up to 225 MBq and 480 MBq, respectively. ⁴³Sc was chemically separated from enriched metallic ⁴⁶Ti (97.0%) and ⁴³CaCO₃ (57.9%) targets, using extraction chromatography. In both cases, ~90% of the final activity was eluted in a small volume of 700 μL, thereby, making it suitable for direct radiolabeling. The prepared products were of high radionuclidic purity, i.e. 98.2% ⁴³Sc were achieved from the irradiation of ⁴⁶Ti, whereas the product isolated from irradiated ⁴³Ca consisted of 66.2% ⁴³Sc and 33.3% ⁴⁴Sc. A PET phantom study performed with ⁴³Sc, via both nuclear reactions, revealed slightly improved resolution over ⁴⁴Sc. In order to assess the chemical purity of the separated ⁴³Sc, radiolabeling experiments were performed with DOTANOC, attaining specific activities of 5–8 MBq/nmol, respectively, with a radiochemical yield of >96%.

Conclusions

It was determined that higher ⁴³Sc activities were accessible via the ⁴³Ca production route, with a comparatively less complex target preparation and separation procedure. The product isolated from irradiated ⁴⁶Ti, however, revealed purer ⁴³Sc with minor radionuclidic impurities. Based on the results obtained herein, the ⁴³Ca route features some advantages (such as higher yields and direct usage of the purchased target material) over the ⁴⁶Ti path when aiming at ⁴³Sc production on a routine basis.

Electronic supplementary material

The online version of this article (10.1186/s41181-017-0033-9) contains supplementary material, which is available to authorized users.

Study on a new design of Tehran Research Reactor for radionuclide production based on fast neutrons using MCNPX code

The aim of this work is to increase the magnitude of the fast neutron flux inside the flux trap where radionuclides are produced. For this purpose, three new designs of the flux trap are proposed and the obtained fast and thermal neutron fluxes compared with each other. The first and second proposed designs were a sealed cube contained air and D₂O, respectively. The results of calculated production yield all indicated the superiority of the latter by a factor of 55% in comparison to the first proposed design. The third proposed design was based on changing the surrounding of the sealed cube by locating two fuel plates near that. In this case, the production yield increased up to 70%.

Practical aspects of 153Gd as a radioactive source for use in brachytherapy

The goal of this study was to investigate the production, purification and immobilization techniques for a ¹⁵³Gd brachytherapy source. We have investigated the maximum attainable specific activity of ¹⁵³Gd through the irradiation of Gd₂O₃ enriched to 30.6% ¹⁵²Gd at McMaster Nuclear Reactor. The advantage of producing ¹⁵³Gd through this production pathway is the possibility to irradiate pre-sealed pellets of ¹⁵²Gd enriched Gd₂O₃, thereby removing the need to perform chemical separation with large quantities of radio-impurities. However, small amounts of long-lived impurities are produced from the irradiation of enriched ¹⁵²Gd targets due to traces of Eu in the sample. If the amount of impurities produced is deemed unacceptable, ¹⁵³Gd can be isolated as an aqueous solution, chemically separated from impurities and loaded onto a sorbent with a high affinity for Gd before encapsulation.

47Sc as useful β--emitter for the radiotheragnostic paradigm: a comparative study of feasible production routes

Background

Radiotheragnostics makes use of the same molecular targeting vectors, labeled either with a diagnostic or therapeutic radionuclide, ideally of the same chemical element. The matched pair of scandium radionuclides, ⁴⁴Sc and ⁴⁷Sc, satisfies the desired physical aspects for PET imaging and radionuclide therapy, respectively. While the production and application of ⁴⁴Sc was extensively studied, ⁴⁷Sc is still in its infancy. The aim of the present study was, therefore, to investigate and compare two different methods of ⁴⁷Sc production, based on the neutron irradiation of enriched ⁴⁶Ca and ⁴⁷Ti targets, respectively.

Methods

⁴⁷Sc was produced by thermal neutron irradiation of enriched ⁴⁶Ca targets via the ⁴⁶Ca(n,γ)⁴⁷Ca → ⁴⁷Sc nuclear reaction and by fast neutron irradiation of ⁴⁷Ti targets via the ⁴⁷Ti(n,p)⁴⁷Sc nuclear reaction, respectively. The product was compared with regard to yield and radionuclidic purity. The chemical separation of ⁴⁷Sc was optimized in order to obtain a product of sufficient quality determined by labeling experiments using DOTANOC. Finally, preclinical SPECT/CT experiments were performed in tumor-bearing mice and compared with the PET image of the ⁴⁴Sc labeled counterpart.

Results

Up to 2 GBq ⁴⁷Sc was produced by thermal neutron irradiation of enriched ⁴⁶Ca targets. The optimized chemical isolation of ⁴⁷Sc from the target material allowed formulation of up to 1.5 GBq ⁴⁷Sc with high radionuclidic purity (>99.99%) in a small volume (~700 μL) useful for labeling purposes. Three consecutive separations were possible by isolating the in-grown ⁴⁷Sc from the ^46/47Ca-containing fraction. ⁴⁷Sc produced by fast neutron irradiated ⁴⁷Ti targets resulted in a reduced radionuclidic purity (99.95–88.5%). The chemical purity of the separated ⁴⁷Sc was determined by radiolabeling experiments using DOTANOC achievable at specific activities of 10 MBq/nmol. In vivo the ⁴⁷Sc-DOTANOC performed equal to ⁴⁴Sc-DOTANOC as determined by nuclear imaging.

Conclusion

The production of ⁴⁷Sc via the ⁴⁶Ca(n,γ)⁴⁷Ca nuclear reaction demonstrated significant advantages over the ⁴⁷Ti production route, as it provided higher quantities of a radionuclidically pure product. The subsequent decay of ⁴⁷Ca enabled the repeated separation of the ⁴⁷Sc daughter nuclide from the ⁴⁷Ca parent nuclide. Based on the results obtained from this work, ⁴⁷Sc shows potential to be produced in suitable quality for clinical application.

Electronic supplementary material

The online version of this article (doi:10.1186/s41181-017-0024-x) contains supplementary material, which is available to authorized users.

Improved GMP-compliant multi-dose production and quality control of 6-[18F]fluoro-L-DOPA

Background

6-[¹⁸F]Fluoro-L-3,4-dihydroxyphenylalanine (FDOPA) is a frequently used radiopharmaceutical for detecting neuroendocrine and brain tumors and for the differential diagnosis of Parkinson’s disease. To meet the demand for FDOPA, a high-yield GMP-compliant production method is required. Therefore, this study aimed to improve the FDOPA production and quality control procedures to enable distribution of the radiopharmaceutical over distances.

FDOPA was prepared by electrophilic fluorination of the trimethylstannyl precursor with [¹⁸F]F₂, produced from [¹⁸O]₂ via the double-shoot approach, leading to FDOPA with higher specific activity as compared to FDOPA which was synthesized, using [¹⁸F]F₂ produced from ²⁰Ne, leading to FDOPA with a lower specific activity. The quality control of the product was performed using a validated UPLC system and compared with quality control with a conventional HPLC system. Impurities were identified using UPLC-MS.

Results

The [¹⁸O]₂ double-shoot radionuclide production method yielded significantly more [¹⁸F]F₂ with less carrier F₂ than the conventional method starting from ²⁰Ne. After adjustment of radiolabeling parameters substantially higher amounts of FDOPA with higher specific activity could be obtained. Quality control by UPLC was much faster and detected more side-products than HPLC. UPLC-MS showed that the most important side-product was FDOPA-quinone, rather than 6-hydroxydopa as suggested by the European Pharmacopoeia.

Conclusion

The production and quality control of FDOPA were significantly improved by introducing the [¹⁸O]₂ double-shoot radionuclide production method, and product analysis by UPLC, respectively. As a result, FDOPA is now routinely available for clinical practice and for distribution over distances.

Production of iodine-124 and its applications in nuclear medicine

Until recently, iodine-124 was not considered to be an attractive isotope for medical applications owing to its complex radioactive decay scheme, which includes several high-energy gamma rays. However, its unique chemical properties, and convenient half-life of 4.2 days indicated it would be only a matter of time for its frequent application to become a reality. The development of new medical imaging techniques, especially improvements in the technology of positron emission tomography (PET), such as the development of new detectors and signal processing electronics, has opened up new prospects for its application. With the increasing use of PET in medical oncology, pharmacokinetics, and drug metabolism, (124)I-labeled radiopharmaceuticals are now becoming one of the most useful tools for PET imaging, and owing to the convenient half-life of I-124, they can be used in PET scanners far away from the radionuclide production site. Thus far, the limited availability of this radionuclide has been an impediment to its wider application in clinical use. For example, sodium [(124)I]-iodide is potentially useful for diagnosis and dosimetry in thyroid disease and [(124)I]-M-iodobenzylguanidine ([(124)I]-MIBG) has enormous potential for use in cardiovascular imaging, diagnosis, and dosimetry of malignant diseases such as neuroblastoma, paraganglioma, pheochromocytoma, and carcinoids. However, despite that potential, both are still not widely used. This is a typical scenario of a rising new star among the new PET tracers.

Automation of (64)Cu production at Turku PET Centre

At Turku PET Centre automation for handling solid targets for the production of (64)Cu has been built. The system consists of a module for moving the target from the irradiation position into a lead transport shield and a robotic-arm assisted setup for moving the target within radiochemistry laboratory. The main motivation for designing automation arises from radiation hygiene.