Terbium-149 production: a focus on yield and quality improvement towards preclinical application.

Terbium-149 (T1/2 = 4.1 h, Eα = 3.98 MeV (16.7%), 28 µm range in tissue) is a radionuclide with potential for targeted alpha therapy. Due to the negligible emission of α-emitting daughter nuclides, toxicity to healthy tissue may be reduced in comparison with other α-particle emitters. In this study, terbium-149 was produced via 1.4 GeV proton irradiation of a tantalum target at the CERN-ISOLDE facility. The spallation products were mass separated and implanted on zinc-coated foils and, later, radiochemically processed. Terbium-149 was separated from the co-produced isobaric radioisotopes and the zinc coating from the implantation foil, using cation-exchange and extraction chromatographic techniques, respectively. At the end of separation, up to 260 MBq terbium-149 were obtained with > 99% radionuclidic purity. Radiolabeling experiments were performed with DOTATATE, achieving 50 MBq/nmol apparent molar activity with radiochemical purity > 99%. The chemical purity was determined by inductively coupled plasma-mass spectrometry measurements, which showed lead, copper, iron and zinc only at ppb level. The radiolabeling of the somatostatin analogue DOTATATE with [149Tb]TbCl3 and the subsequent in vivo PET/CT scans conducted in xenografted mice, showing good tumor uptake, further demonstrated product quality and its ability to be used in a preclinical setting.

Optimization of 68Ga production at an 18 MeV medical cyclotron with solid targets by means of cross-section measurement of 66Ga, 67Ga and 68Ga.

The future development of personalized nuclear medicine relies on the availability of novel medical radionuclides. In particular, radiometals are attracting considerable interest since they can be used to label both proteins and peptides. Among them, the ß+-emitter 68Ga is widely used in nuclear medicine for positron emission tomography (PET). It is used in theranostics as the diagnostic partner of the therapeutic ß--emitters 177Lu and 90Y for the treatment of a wide range of diseases, including prostate cancer. Currently, 68Ga is usually obtained via 68Ge/68Ga generators. However, their availability, high price and limited produced radioactivity per elution are a major barrier for a wider use of the 68Ga-based diagnostic radiotracers. A promising solution is the production of 68Ga by means of proton irradiation of enriched 68Zn liquid or solid targets. Along this line, a research program is ongoing at the Bern medical cyclotron, equipped with a solid target station. In this paper, we report on the measurements of 68Ga, 67Ga and 66Ga production cross-sections using natural Zn and enriched 68Zn material, which served as the basis to perform optimized 68Ga production tests with enriched 68Zn solid targets.

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

BACKGROUND: Terbium-155 [T1/2 = 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.

Large-scale production of 88Y and 88Zr/88Y generators: A proof of concept study for a 70 MeV H- cyclotron.

The large-scale production of 88Y with proton-induced reactions has been investigated from the perspective of new generation 70 MeV H- cyclotrons. Tandem target configurations are presented for both the direct production of 88Y as well as for producing 88Zr/88Y generators. Based on the relevant excitation functions, physical yields have been derived for 88Y production with Y2O3/SrCO3 tandem targets and 88Zr production with Zr/Y2O3 tandem targets. Yields are presented for optimized targets (i.e. optimum yield) as well as for balanced thermal loads on the individual targets. Liquid 88Zr/88Y generators have been produced using both natural Zr and Nb target materials, the former for dedicated productions and the latter as a byproduct by processing spent irradiated Nb capsules which normally would constitute radioactive waste. These stock solutions, which contain both the target material and 88Zr precursor, are retained virtually unchanged after processing except for the removal of 88Y on AG MP-50 macroporous cation-exchange resin. Methods are presented for the preparation of Nb stock solutions in hydrofluoric acid and Zr stock solutions in sulphuric acid. It is shown that multi-Ci productions of 88Y are feasible at a 70 MeV cyclotron facility, suitable for the needs of fracking applications. In addition, 88Zr/88Y generators can provide 88Y with very high specific activity, suitable for labelling of biomolecules. LA-UR-20-24305.

The Metamorphosis of Radionuclide Production and Development at Paul Scherrer Institute.

Radionuclide production and development has a long history at the Paul Scherrer Institute (PSI) and dates back to the founding times of its forerunner institutions: the Federal Institute for Reactor Research and the Swiss Institute for Nuclear Research. The facilities used for this purpose have evolved substantially over the last five decades. Many radiometals in use today, as radiopharmaceuticals, are for the diagnosis and treatment of disease, with the most popular means of detection being Positron Emission Tomography. These positron emitters are easily produced at low proton energies using medical cyclotrons, however, developments at these facilities are lacking. Currently, the fixed 72 MeV proton beam at PSI is degraded at IP2 irradiation station to provide the desired energy to irradiate targets to produce the likes of 44Sc, 43Sc and 64Cu as a proof of principle, which are of great interest to the nuclear medicine community. This development work can then be implemented at facilities containing medical cyclotrons. A history of the development of radionuclides at PSI, along with current development and projects with partner institutions, is described.

Establishment of a clinical SPECT/CT protocol for imaging of 161Tb.

BACKGROUND: It has been proposed, and preclinically demonstrated, that 161Tb is a better alternative to 177Lu for the treatment of small prostate cancer lesions due to its high emission of low-energy electrons. 161Tb also emits photons suitable for single-photon emission computed tomography (SPECT) imaging. This study aims to establish a SPECT protocol for 161Tb imaging in the clinic. MATERIALS AND METHODS: Optimal settings using various γ-camera collimators and energy windows were explored by imaging a Jaszczak phantom, including hollow-sphere inserts, filled with 161Tb. The collimators examined were extended low-energy general purpose (ELEGP), medium-energy general purpose (MEGP), and low-energy high resolution (LEHR), respectively. In addition, three ordered subset expectation maximization (OSEM) algorithms were investigated: attenuation-corrected OSEM (A-OSEM); attenuation and dual- or triple-energy window scatter-corrected OSEM (AS-OSEM); and attenuation, scatter, and collimator-detector response-corrected OSEM (ASC-OSEM), where the latter utilized Monte Carlo-based reconstruction. Uniformity corrections, using intrinsic and extrinsic correction maps, were also investigated. Image quality was assessed by estimated recovery coefficients (RC), noise, and signal-to-noise ratio (SNR). Sensitivity was determined using a circular flat phantom. RESULTS: The best RC and SNR were obtained at an energy window between 67.1 and 82.1 keV. Ring artifacts, caused by non-uniformity, were removed with extrinsic uniformity correction for the energy window between 67.1 and 82.1 keV, but not with intrinsic correction. Analyzing the lower energy window between 48.9 and 62.9 keV, the ring artifacts remained after uniformity corrections. The recovery was similar for the different collimators when using a specific OSEM reconstruction. Recovery and SNR were highest for ASC-OSEM, followed by AS-OSEM and A-OSEM. When using the optimized parameter setting, the resolution of 161Tb was higher than for 177Lu (8.4 ± 0.7 vs. 10.4 ± 0.6 mm, respectively). The sensitivities for 161Tb and 177Lu were 7.41 and 8.46 cps/MBq, respectively. CONCLUSION: SPECT with high resolution is feasible with 161Tb; however, extrinsic uniformity correction is recommended to avoid ring artifacts. The LEHR collimator was the best choice of the three tested to obtain a high-resolution image. Due to the complex emission spectrum of low-energy photons, window-based scatter correction had a minor impact on the image quality compared to using attenuation correction only. On the other hand, performing attenuation, scatter, and collimator-detector correction clearly improved image quality. Based on these data, SPECT-based dosimetry for 161Tb-labeled radiopharmaceuticals is feasible.

Extending the life of SnO268Ge/68Ga generators used in the radiolabelling of ion exchange resins.

The SnO268Ge/68Ga generator system is widely used in medical imaging to provide a regular supply of the radionuclide 68Ga (T½ = 68.3 min) for positron emission tomography (PET). These generators are also used to supply 68Ga for the fabrication of tracer particles for application in positron emission particle tracking (PEPT). The tracer particles are fabricated by radiolabelling ion exchange resins such as Purolite NRW100 with 68Ga; however, contaminants from the degradation of the SnO2 column over time interfere with the uptake of 68Ga. The major contaminants are Zn(II), Fe(III) and Sn(IV) with 68Ge (IV) being eluted from the column as it degrades. This paper describes an improved method to purify the 68Ga supply using an Amberchrom CG-71m absorption resin column integrated into a newly designed separation panel. This method reduces the amount of Zn(II) and Fe(III) in the 68Ga eluate and improves the radiolabelling performance by more than 10% when compared to the un-purified product. The method can extend the life-span of the generator by several months.

Concurrent spectrometry of annihilation radiation and characteristic gamma-rays for activity assessment of selected positron emitters.

A method is described to determine the activity of non-pure positron emitters in a radionuclide production environment by assessing the 511keV annihilation radiation concurrently with selected γ-lines, using a single High-Purity Germanium (HPGe) detector. Liquid sources of 22Na, 52Fe, 52mMn, 61Cu, 64Cu, 65Zn, 66Ga, 68Ga, 82Rb, 88Y, 89Zr and 132Cs were prepared specifically for this study. Acrylic absorbers surrounding the sources ensured that the emitted ß+-particles could not escape and annihilate away from the source region. The absorber thickness was matched to the maximum ß+ energy for each radionuclide. The effect on the 511keV detection efficiency by the non-homogeneous distribution of annihilation sites inside the source and absorber materials was investigated by means of Monte Carlo simulations. It was found that no self-absorption corrections other than those implicit to the detector calibration procedure needed to be applied. The medically important radionuclide, 64Cu, is of particular interest as its strongest characteristic γ-ray has an intensity of less than 0.5%. In spite of the weakness of its emission intensity, the 1346keV γ-line is shown to be suitable for quantifying the 64Cu production yield after chemical separation from the target matrix has been performed.

Production of (28)Mg by bombardment of (nat)Cl with 200MeV protons: Proof-of-concept study for a stacked LiCl target.

A stacked target consisting of ten Al-encapsulated LiCl discs, for producing (28)Mg via the (nat)Cl(p,X)(28)Mg process in the energy region 50-200MeV, is described. This target was irradiated with a 200MeV beam at an intensity of 100nA, providing information on both yield and outscattering losses. Results of a Monte Carlo modelling of the beam and target, by means of the code MCNPX, are also presented. Similar Al-encapsulated LiCl discs were individually irradiated with 66MeV proton beams of 65 and 90µA, respectively, to study their behaviour under high-intensity bombardment. Once removed from the Al encapsulation, the (28)Mg can be separated from the LiCl target material efficiently, using a 12.5cm x 1cm(2) column containing Purolite S950 chelating resin. The eluate contains (7)Be but no other measurable radio-contaminants. The removal of the (7)Be contaminant is performed by cation exchange chromatography in malate media, with (28)Mg being retained by the resin and (7)Be eluted.

The use of selective volatization in the separation of 68Ge from irradiated Ga targets.

Cyclotron-produced (68)Ge can be separated from its Ga target material by dissolving the target in aqua regia and collecting the volatile (68)Ge in a solution containing 1.0M NaOH and 2% Na2SO3. The solution is then acidified with HF before being loaded onto a column containing AG MP-1 anion exchange resin. The column is rinsed with dilute HF to remove any remaining impurities, before eluting the desired product with 0.1M HCl. A radiochemically pure product is obtained.

The production of (88)Y in the proton bombardment of (nat)Sr: New excitation and separation studies.

The cyclotron production of (88)Y at iThemba LABS is performed via the reaction (88)Sr(p,n)(88)Y. The yields obtained were inconsistent with nuclear data obtained from the literature and the excitation function of the nuclear reaction was re-measured, using a differentiation of thick-target production rate measurements. Ion exchange chromatographic methods are described to separate (88)Y from (nat)Sr target material using AG MP-1 resin and AG 50W-X4 resins, respectively.