Radionuclide Generator Research Articles

Gas-chemical approach to 225Ac/213Bi radionuclide generator for medical application

Gas-chemical approach to 225Ac/213Bi radionuclide generator for medical application

Study on the sorption and desorption behavior of La3+ and Bi3+ by bis(2-ethylhexyl)phosphate modified activated carbon.

The separation of 213Bi from its parent radionuclide 225Ac via radionuclide generators has proven to be a challenge due to the limited performance of the current sorbents. This study evaluated the separation performance of La3+ (as a surrogate for 225Ac) and Bi3+ using bis(2-ethylhexyl)phosphate monofunctionalized activated carbon (HDEHP/AC). The potential applications of phosphate groups as active sites and the carbon structure as a sorbent support were confirmed and validated. Various factors, including pH values, salt concentration, halide ions, contact time, solid-to-liquid ratio, initial La3+/Bi3+ concentration, and gamma irradiation were examined through batch sorption experiments in both single and binary systems. HDEHP/AC had a high sorption capacity for La3+ via electrostatic attraction, with the sorption data fitting well to the pseudo-second-order kinetic equation and Langmuir model. The sorption performance of La3+ on HDEHP/AC was minimally affected as the NaCl/NaI concentrations increased at pH = 2, whereas the sorption capacity for Bi3+ decreased significantly. Additionally, selective desorption of La3+ and Bi3+ was achieved using HNO3 and NaI solutions, respectively. These results backed up by a conceptual separation process point toward a potential use of these materials in a direct/inverse 225Ac/213Bi radionuclide generator. Further optimization of the material and separation process will be required to bring this class of promising materials into an actual generator for medical applications.

Automated Production of [68Ga]Ga-Desferrioxamine B on Two Different Synthesis Platforms.

Background/Objectives: PET imaging of bacterial infection could potentially provide added benefits for patient care through non-invasive means. [68Ga]Ga-desferrioxamine B-a radiolabelled siderophore-shows specific uptake by human-pathogenic bacteria like Staphylococcus aureus or Pseudomonas aeruginosa and sufficient serum stability for clinical application. In this report, we present data for automated production of [68Ga]Ga-desferrioxamine B on two different cassette-based synthesis modules (Modular-Lab PharmTracer and GRP 3V) utilising commercially obtainable cassettes together with a licensed 68Ge/68Ga radionuclide generator. Methods: Quality control, including the determination of radiochemical purity, as well as a system suitability test, was set up via RP-HPLC on a C18 column. The two described production processes use an acetic acid/acetate buffer system with ascorbic acid as a radical scavenger for radiolabelling, yielding ready-to-use formulations with sufficient activity yield. Results: Batch data analysis demonstrated radiochemical purity of >95% by RP-HPLC combined with ITLC and excellent stability up to 2 h after synthesis. Specifications for routine production were set up and validated with four masterbatches for each synthesis module. Conclusions: Based on this study, an academic clinical trial for imaging of bacterial infection was initiated. Both described synthesis methods enable automated production of [68Ga]Ga-desferrioxamine B in-house with high reproducibility for clinical application.

Development of a SnO2-based 44Ti/44Sc generator for medical applications

Towards application of 44Sc for diagnostic nuclear medicine, a 44Ti/44Sc generator based on an inorganic resin has been evaluated. Unlike other radionuclide generators used for medical applications, the long-term retention of the parent 44Ti is vital due to its long half life. Herein, tin dioxide (SnO2), a robust inorganic-based resin, has been synthesized and used as the stationary phase for a 44Ti/44Sc generator. The sorption behavior of 44Ti/44Sc was tested on SnO2 with varying acids, concentrations, and times. Preliminary batch study results showed >88 % 44Ti retention to the resin at lower acid concentrations (0.05 M HNO3 and 0.05 M HCl). A pilot generator was evaluated for a year, demonstrating 85.3 ± 2.8 % 44Sc elution yields and 0.71 ± 0.14 % 44Ti breakthrough in 5 M HNO3. Based on capacity studies, a 7.4 MBq (200 µCi) upscaled generator system was constructed for further evaluation of the SnO2 resin stability and the efficacy of the eluted 44Sc for radiolabeling. 44Sc could be regularly eluted from this generator in 5 M HNO3 with an overall average radiochemical yield 84.7 ± 9.5 %. Post-elution processing of the 44Sc with DGA-normal resin removed all 44Ti present and allowed for high 44Sc-DOTA labeling yields of 94.2 ± 0.5 %. Overall, SnO2 has been shown to be a viable material for a 44Ti/44Sc generator.

Feasibility of producing 225Ac via thermal neutron irradiation of 226Ra: A systematic theoretical study

With a suitable half-life and abundant radiolabeling strategy, 225Ac has become one of the most promising radionuclides in the area of targeted alpha therapy. However, limited radionuclide supply is threatening the development of 225Ac related endoradiotherapy dramatically. As the parent nuclide of 225Ac, 229Th can be produced via 226Ra(3n, 2β)229Th reaction in a nuclear reactor. However, related practice has not been conducted in large scale, since the nuclear reaction pathway for producing 229Th is complicated. In this work, the feasibility of producing 225Ac/229Th in a reactor was confirmed by systematic theoretical calculations, and a procedure that combines irradiation with separation process was proposed. The results show that 176 MBq of 229Th can be produced by irradiating 1.0 g of 226Ra with a neutron flux density of 1 × 1015 n cm−2 s−1 for 90 days. This will generate 150 MBq of 225Ac monthly from a radionuclide generator, which is sufficient for the single treatment cycle of 200 patients each year considering the radioactivity loss in radiochemical separation, transfer and radiolabeling process. In addition, this irradiation process will also produce 37.8 GBq 227Ac for the preparation of 227Ac-227Th-223Ra generator. In general, the production of 225Ac by neutron irradiation of 226Ra in reactor is practicable and holds potential to alleviate the shortage of current supply of 225Ac.

An electrochemical generator for the continual supply of 213Bi from 225Ac for use in targeted alpha therapy applications

Bismuth-213 is a radionuclide of interest for targeted alpha therapy and is supplied via a radiochemical generator system through the decay of 225Ac. Radionuclide generators employ longer lived "parent" radionuclides to routinely supply shorter-lived "daughter" radionuclides. The traditional 225Ac/213Bi radiochemical generator relies on an organic cation exchange resin where 225Ac binds to the resin and 213Bi is routinely eluted. These resins degrade when they absorb large doses of ionizing radiation (>1×106Gy/mg), which has been observed when the loading activity of 225Ac exceeds 2.59*109Bq (70mCi). Herein we report the development of an electrochemical generator for the supply of 213Bi that has the potential to overcome this limitation. Bismuth-213 spontaneously electrodeposits onto nickel foils in 0.1M hydrochloric acid at 70°C. Using this method, we were able to plate an average of 73±4% of the 213Bi in solution and obtain a final 213Bi recovery of 65±8% in 0.1M citrate pH4.5 via reverse electrolysis using titanium as the cathode. The recovered 213Bi had an average radiochemical purity of >99.8% and was successfully used to radiolabel DOTATATE with an average radiochemical yield of 85.1% (not optimized).

68Ge/68Ga composite CeO2-PAN generator: preparation, testing and use

This work is focused on the testing of a new 68Ge/68Ga radionuclide generator based on cerium dioxide in polyacrylonitrile beads. During an 18-month period, parameters of elution were monitored, and basic radiolabelling studies were carried out. As current commercial solutions deal with high elution volume and low long-term stability, the constructed system offers several improvements. The composite sorbent provides uniformity of active component dispersion and high capacity for 68Ge. It enables minimal elution volume without fractionation (up to 1.8 mL) and stability of 68Ge breakthrough (under 0.001%) after initial period of wash-out and low cerium leakage (under 5 ppm).

Preparation of Medical 228Th-224Ra Radionuclide Generator Based on SiO2@TiO2 Microspheres.

224Ra (T1/2 = 3.63 d), an α-emitting radionuclide, holds significant promise in cancer endoradiotherapy. Current 224Ra-related therapy is still scarce because of the lack of reliable radionuclide supply. The 228Th-224Ra radionuclide generator can undoubtedly introduce continuous and sustainable availability of 224Ra for advanced nuclear medicine. However, conventional metal oxides for such radionuclide generators manifest suboptimal adsorption capacities for the parent nuclide, primarily attributable to their limited surface area. In this work, core-shell SiO2@TiO2 microspheres were proposed to develop as column materials for the construction of a 228Th-224Ra generator. SiO2@TiO2 microspheres were well prepared and systematically characterized, which has also been demonstrated to have good adsorption capacity to 228Th and very weak binding affinity toward 224Ra via simulated chemical separation. Upon introducing 228Th-containing solution onto the SiO2@TiO2 functional column, a 228Th-224Ra generator with excellent retention of the parent radionuclide and ideal elution efficiency of daughter radionuclide was obtained. The prepared 228Th-224Ra generator can produce 224Ra with high purity and medical usability in good elution efficiency (98.72%) even over five cycles. To the best of our knowledge, this is the first time that the core-shell mesoporous materials have been applied in a radionuclide generator, which can offer valuable insights for materials chemistry, radiochemical separation, and biological medicine.

An Analysis of the Diagnostic Performance of Tc-99m PSMA SSPECT/CT in Biochemically Recurrent Prostate Cancer Compared with Ga-68 PSMA PET/CT: A Single-center, Prospective Study.

Biochemical recurrence (BCR) after initial management of Prostate Carcinoma (PC) is frequent. Subsequent interventions rely on disease burden and metastasis distribution. 68Ga prostate-specific membrane antigen positron emission tomography/computed tomography (PSMA PET/CT) is an excellent imaging modality in BCR. However, 68Ga is radionuclide generator produced and has restricted availability. 99mTc-labeled PSMA could be a potential cost-effective alternative. We compared the performance of 99mTc-PSMA single-photon emission CT (SPECT)/CT and 68Ga-PSMA PET/CT in BCR with a serum prostate surface antigen (PSA) level of <20 ng/mL. The prospective study included 25 patients with BCR and at least one lesion on a 68Ga-PSMA PET/CT. All patients underwent 99 mTc-PSMA SPECT/CT, and disease distribution and metastatic burden were compared with 68Ga-PSMA PET/CT. The maximum standard uptake value (SUVmax) and the tumor-to-background ratio (TBR) were computed and analyzed. The mean age and serum PSA (SPSA) were 69.72 ± 6.69 years and 5.65 ± 6.07 ng/mL. Eleven patients (44%) had SPSA ≤2 ng/mL. Recurrent sites were noted in the prostate (19, 76%), prostatic bed (3, 12%), and pelvis lymph nodes (LNs) (13, 52%). Distant metastasis to bones (13, 52%), lungs (5, 20%), and retroperitoneal LNs (2, 8%) were noted. Both modalities were concordant for the recurrent disease at the prostate, prostatic bed, bone, and lung lesions. 99mTc-PSMA could localize pelvis LNs in most patients (10/13, 76.9%). The site-specific sensitivity and specificity between the two modalities were not significantly different (P > 0.05). TBR shows excellent correlation with SUVmax (0.783, P < 0.001). Four (16%) patients were understaged with 99mTc-PSMA due to the nonvisualization of the subcentimeter size LNs. No patient with systemic metastases was understaged. 99mTc-PSMA SPECT/CT has good concordance with 68Ga-PSMA PET/CT in BCR, even at low PSA levels. However, it may miss a few subcentimeter LNs due to lower resolution. 99mTc-PSMA SPECT/CT could be a simple, cost-effective, and readily available imaging alternative to PET/CT.

Development of 225Ac/213Bi generator based on α-ZrP-PAN composite for targeted alpha therapy

BackgroundRadioligand therapy using alpha emitters has gained more and more prominence in the last decade. Despite continued efforts to identify new appropriate radionuclides, the combination of 225Ac/213Bi remains among the most promising. Bismuth-213 has been employed in clinical trials in combination with appropriate vectors to treat patients with various forms of cancer, such as leukaemia, bladder cancer, neuroendocrine tumours, melanomas, gliomas, or lymphomas. However, the half-life of 213Bi (T½ = 46 min) implies that its availability for clinical use is limited to hospitals possessing a 225Ac/213Bi radionuclide generator, which is still predominantly scarce. We investigated a new Ac/Bi generator system based on using the composite sorbent α-ZrP-PAN (zirconium(IV) phosphate as active component and polyacrylonitrile as matrix). The developed 225Ac/213Bi generator was subjected to long-term testing after its development. The elution profile was determined and the elution yield, the contamination of the eluate with the parent 225Ac and the contamination of the eluate with the column material were monitored over time. ResultsThe high activity (75 MBq of parent 225Ac) generator with a length of 75 mm and a diameter of 4 mm containing the composite sorbent α-ZrP-PAN with a particle size of 0.8 to 1.0 mm as the stationary phase, eluted with a mixture of 10 mM DTPA in 5 mM nitric acid, provided 213Bi with yields ranging from 77 % to 96 % in 2.8 mL of eluate, with parent 225Ac contamination in the order of 10−3 %, up to twenty days of use. ConclusionAll the results of the monitored parameters indicate that the composite sorbent α-ZrP-PAN based separation system for the elution of 213Bi is a very promising and functional solution.

Gallium and germanium static and kinetic sorption and desorption studies on cerium dioxide nanoparticles

Positron-emitting 68Ga has become an integral part of nuclear medicine. It is a radionuclide obtained from 68Ge/68Ga radionuclide generators. Apart from other qualities, it is a feature that have caused its growing popularity in medical practice. Nevertheless, the increasing demand for 68Ga supply results in the need of new and improved separation systems to obtain it. Hydrous oxides such as titanium or zirconium dioxide are usually the preferred separation materials, but other oxidic materials are under study. Thorough testing of the sorption behaviour is often overlooked in these studies, though it is indispensable for the applicability and suitability assessment. This work is focused on the sorption behaviour description of cerium dioxide prepared by calcination of ceric nitrate. Both batch and kinetic sorption experiments were conducted. Such a thorough study has never been carried out for this material. In batch experiments, the optimal conditions for separation of 68Ga and 68Ge were found. Kinetic experimental data were used for mathematical modelling. Six kinetic models derived from various rate-controlling processes were used and their applicability was demonstrated. The diffusion in an inert layer is the rate-controlling process of both gallium and germanium sorption and desorption. This model description of sorption kinetics will further enable us to optimise the sorption and desorption processes numerically. Thus far, it was found out that the prepared cerium dioxide can quantitatively adsorb germanium in very short time of 10 min from 1mM hydrochloric acid and the most convenient medium for 68Ge and 68Ga separation is 0.1 M hydrochloric acid.

P-314 - Nanoparticles as a Radionuclide Generator System for 225Ac/213Bi

P-314 - Nanoparticles as a Radionuclide Generator System for 225Ac/213Bi

Separation of 44mSc/44gSc Nuclear Isomers Based on After-Effects.

44gSc presents a particular interest for application in nuclear medicine for positron emission tomography (PET) due to its favorable nuclear decay properties (t1/2 = 3.97 h, Emax = 1.47 MeV, branching ratio 94.3% β+). Its nuclear isomer 44mSc (t1/2 = 58.61 h) decays by isomeric transition (IT) into 44gSc, accompanied by ≈12% of conversion electron emission, which can cause a partial release of the daughter 44gSc from the chelate complex. A 13 MeV cyclotron at TRIUMF was used to produce both 44mSc and 44gSc via the natCa(p,n)44m,gSc reaction. A 44mSc/44gSc generator was designed by using a Strata C-18E cartridge. After several tested systems, a successful separation method was developed using DOTATOC as a chelator, a Strata C-18E cartridge as a generator column, and an elution solution of 0.1 M NH4-α-HIB. The yield of the generator with the daughter 44gSc release was equal to 9.8 ± 1.0% (or ≈80% per portion of conversion). This result shows the important role of after-effects in the design of radionuclide generators. Nuclear cross-section calculations were applied using the TALYS code to allow for the determination of the most promising alternative routes for 44mSc production, which will enable the development of a full-scale 44mSc/44gSc radionuclide generator based on after-effects.

Automated production of [68Ga]Ga-DOTA-exendin-4 via fractionated radionuclide generator elution on a cassette based synthesis module

BackgroundPET/CT imaging of glucagon-like peptide receptor 1 has recently filled a gap in reliably diagnosing insulinoma through non-invasive means. 68Ga-labelled derivatives of exendin-4 show high sensitivity as well as sufficient serum stability to enable routine clinical application. Here, we provide data for automated production of [68Ga][Nle14,Lys40(Ahx-DOTA-Ga)NH2]exendin-4 ([68Ga]Ga-DOTA-exendin-4) on a cassette based synthesis module (Modular-Lab PharmTracer, Eckert & Ziegler) using commercially available cassettes in combination with an approved 68Ge/68Ga generator (GalliaPharm, Eckert & Ziegler). This setup ensured high reproducibility as well as low radiation burden for the production team. Quality control including determination of radiochemical purity was performed by RP-HPLC using a water/0.1 % TFA/acetonitrile gradient on a C18 column. A modified TLC system with ammonium acetate & methanol as mobile phase and a novel limit test for determination of polysorbate 80 content in the final formulation are also described in this study. Main findingsReliable yields as well as high molar activity for patient use were only achieved using a fractionated elution approach. Batch data showed radiochemical purity of >93 % as determined by RP-HPLC and TLC as well as good stability over 2 h post production. Testing for polysorbate 80 confirmed a concentration <1 mg/mL in the final product solution. Specifications for routine production were established based on existing Pharmacopeia monographs for other radiopharmaceuticals and were validated with 5 master batches. ConclusionThe described synthesis method enables reproducible, automated in-house production of [68Ga]Ga-DOTA-exendin-4 for routine clinical application.

Selective separation of Bi3+ from La3+/Ac3+ by sorption on sulfonated carbon materials for use in an inverse 225Ac/213Bi radionuclide generator: Batch and column tests

213Bi-labeled radiopharmaceuticals for cancer treatment are receiving considerable interest. 213Bi must be separated from the parent radionuclide 225Ac in a radionuclide generator, but the choice of a suitable sorbent material with high chemical and radiolytic stability is challenging. Present work reports on the synthesis of sulfonated carbon materials (SCMs) prepared by carbonization of methyl cellulose at temperatures between 300 and 800 °C, followed by a sulfonation process. These SCMs were evaluated as sorbents in inverse 225Ac/213Bi generators. The sulfonated carbon material prepared at 500 °C (SCM-500) was selected as the most promising candidate sorbent based on its unique aromatic carbon structure, sufficient amount of oxygen-containing functional groups, and selective sorption performance towards La3+/Bi3+. In addition, various variables, including the acidity of the solutions, salt concentrations, and sorption time have been evaluated for separating Bi3+ from La3+ (as a surrogate of Ac3+) by performing batch sorption experiments. SCM-500 was found to exhibit a high sorption Bi3+ selectivity over La3+ at low pH or in the presence of a high concentration of NaNO3. The optimized conditions in a column tests with this sorbent demonstrated a high chemical 213Bi yield (94 ± 3%) with less than 0.04% 225Ac relative to the eluted 213Bi activity. These results indicate that SCM-500 is a promising material for use in an inverse 225Ac/213Bi radionuclide generator. Furthermore, a conceptual design of an inverse 225Ac/213Bi radionuclide for medical applications is suggested based on the separation properties of SCM-500 and AG MP-50.

Rapid Elution of 226Th from a Two-Column 230U/226Th Generator with Diluted and Buffer Solutions.

A radionuclide generator of the short-lived alpha emitter 226Th was proposed. An original scheme consisting of two in-series chromatographic columns was developed for rapidly producing a neutral citric-buffered eluate of high purity 226Th. The first column filled with TEVA resin retained the parent 230U, while 226Th was eluted with 7 M HCl solution to be immediately adsorbed on the second column containing DGA resin or UTEVA resin. Having substituted the strongly acidic medium of second column with neutral salt solution, 226Th was desorbed with diluted citric buffer solution. One cycle of generator milking took 5-7 min and produced >90% of 226Th in 1.5 mL of eluate (pH 4.5-5.0) appropriate for direct use in radiopharmaceutical synthesis. The 230U impurity in 226Th eluate was less than 0.01%. The proposed two-column 230U/226Th generator was tested over 2 months including a second loading of 230U additionally accumulated from 230Pa.

Sorption and desorption performance of La3+/Bi3+ by surface-modified activated carbon for potential application in medical 225Ac/213Bi generators

213Bi is a promising candidate for targeted alpha therapy.However, the search for suitable materials to separate clinically relevant activities of 213Bi from its parent 225Ac remains an ongoing yet critical challenge. This study evaluates sulfonated Norit CA1 (activated carbon) materials (SNCAs) as alternative sorbents for use in 225Ac/213Bi radionuclide generators and gives an overview of their simultaneous sorption/desorption performance towards La3+ and Bi3+. The SNCAs exhibit high and selective uptake of Bi3+ in the presence of NaNO3 at low pH values. In contrast, the sorption performance for La3+ (which was certified as a non-radioactive surrogate for 225Ac3+) onto SNCAs was more sensitive to the salt concentrations and pH values. The desorption results show that an efficient elution of Bi3+ from SNCAs is achieved using HCl, NaI or NaCl solutions. The experimental sorption data are analysed by various kinetic and isotherm models. The SNCAs exhibit high radiolytic stability, fast sorption kinetics, good durability and recyclability for La3+/Bi3+. This research indicates that SNCAs with sufficient oxygen-containing groups are promising sorbents for the separation of 213Bi in an inverse 225Ac/213Bi radionuclide generator.

Preparation and Surface Characterization of Cerium Dioxide for Separation of 68Ge/68Ga and Other Medicinal Radionuclides.

The overall need for the preparation of new medicinal radionuclides has led to the fast development of new sorption materials, extraction agents, and separation methods. Inorganic ion exchangers, mainly hydrous oxides, are the most widely used materials for the separation of medicinal radionuclides. One of the materials that has been studied for a long time is cerium dioxide, a competitive sorption material for the broadly used titanium dioxide. In this study, cerium dioxide was prepared through calcination of ceric nitrate and fully characterized using X-ray powder diffraction (XRPD), infrared spectrometry (FT-IR), scanning and transmission electron microscopy (SEM and TEM), thermogravimetric and differential thermal analysis (TG and DTA), dynamic light scattering (DLS), and analysis of surface area. In order to estimate the sorption mechanism and capacity of the prepared material, characterization of surface functional groups was carried out using acid-base titration and mathematical modeling. Subsequently, the sorption capacity of the prepared material for germanium was measured. It can be stated that the prepared material is prone to exchange anionic species in a wider range of pH than titanium dioxide. This characteristic makes the material superior as a matrix in 68Ge/68Ga radionuclide generators, and its suitability should be further studied in batch, kinetic, and column experiments.

Gamma radiation effects on AG MP-50 cation exchange resin and sulfonated activated carbon for bismuth-213 separation.

Medical 225Ac/213Bi radionuclide generators are designed to provide a local supply of the short-lived 213Bi for cancer treatment. However, radiation-induced damage to the sorbents commonly used in such radionuclide generators remains a major concern. In this study, the effects of gamma radiation on AG MP-50 cation exchange resin and sulfonated activated carbon (SAC) were studied by analyzing the changes in the morphological characteristics, functional groups, and the La3+/Bi3+ sorption performance, with La3+ being a suitable non-radioactive substitute for Ac3+. The surface sulfonic acid groups of AG MP-50 resin suffered from severe radiation-induced degradation, while the particle morphology was changed markedly after being exposed to absorbed doses of approximately 11 MGy. As a result, the sorption performance of irradiated AG MP-50 for La3+ and Bi3+ was significantly decreased with increasing absorbed doses. In contrast, no apparent changes in acquired morphological characteristics were observed for pristine and irradiated SAC based on SEM and XRD characterization. The surface oxygen content (e.g., O-C[double bond, length as m-dash]O) of irradiated SAC increased for an absorbed dose of 11 MGy due to free radical-induced oxidation. The sorption performance of pristine and irradiated SAC materials for La3+ and Bi3+ remained generally the same at pH values of 1 and 2. Furthermore, the applicability of AG MP-50 and SAC in the 225Ac/213Bi generators was illustrated in terms of their radiolytic stability. This study provides further evidence for the practical implementation of both AG MP-50 and SAC in 225Ac/213Bi radionuclide generators.

Evaluation of hydroxamate-based resins towards a more clinically viable 44Ti/44Sc radionuclide generator

Several hydroxamate-based resins were synthesized and tested for use in 44Ti/44Sc generator systems in small scale experiments (740 kBq 44Ti). The most promising resin was tested further in larger scale generator studies (37 MBq). This resin displayed impressive retention of 44Ti over several elutions, and high quantities of 44Sc were obtained in small volumes of dilute HCl eluents. Initial radiolabeling experiments were conducted and demonstrated the possibility of direct radiolabeling of the generator produced 44Sc with DOTA.