Radionuclide Generator Research Articles

Phase and structural investigations of the content of natural radionuclides in rock and inorganic building materials

The population living in areas rich in granites, minerals and soils with high concentrations of natural radionuclides is exposed to higher external radiation. Significant part of a person's time resides in buildings where he lives and works. The walls, floors and roofs of the buildings shield (weaken) the external gamma background. The aim of the work is to determine the content of natural radionuclides in rock and inorganic materials used as building materials of samples taken from different buildings in Southern Bulgaria by analysing the samples and measuring their natural radioactivity, then to compare the results with the maximum concentrations of activity in common building materials. also used in the EU.The data obtained from the analyzes do not give grounds to expect the generation of radionuclides from natural and conventional inorganic building materials used in the studied areas.

How Efficient Are Monte Carlo Calculations Together With the Q-System to Determine Radioactive Transport Limits? Case Study on Medical Radionuclides.

The development of the so-called theranostics approach, in which imaging information are used to define a personalized therapeutic strategy, is driving the increasing use of radionuclides in nuclear medicine. They are artificially produced either in nuclear reactors, charged particle accelerators, or using radionuclide generators. Each method leads to radioisotopes with different characteristics and then clinical utility. In the first two cases they are extracted from stable or radioactive target bombarded with a particle beam. After extraction/purification of the target, the radionuclides, either implanted on solid or in liquid form, needs to be transported to a centralized production site, a radiopharmacy or an hospital. The transport of needed radioactive material must obey strict rules. For a radionuclide, a limit in activity that it is possible to transport has been established for each type of allowed packages. For type A package these limits are called A1 (for special form sources, i.e., certified perfectly sealed and encapsulated sources) and A2 (for non-special form sources). However, these limits can be easily reached if the activity to transport is high or if the radionuclide of interest is a “non–conventional” one. Indeed, for many radionuclides, there are no available/tabulated A1 and A2 and, in these cases, a very conservative set of values is imposed. This is in particular the case for some of the non-conventional radionuclide of interest in medicine (as for example Tb-149 or Tb-161). The non-tabulated values, and in general the A1/A2 limit, can be evaluated following the so-called Q-system and using Monte Carlo calculations. In the present work, we have used the MCNPX Monte Carlo code to evaluate dose rate values in different exposure scenarios. This has allowed us to determine A1/A2 coefficients for several non-conventional radionuclides of interest for medical applications. The developed technique can be extended easily to other radionuclides and can be adapted in case of changes in regulatory rules.

Nuclear waste from small modular reactors

SignificanceSmall modular reactors (SMRs), proposed as the future of nuclear energy, have purported cost and safety advantages over existing gigawatt-scale light water reactors (LWRs). However, few studies have assessed the implications of SMRs for the back end of the nuclear fuel cycle. The low-, intermediate-, and high-level waste stream characterization presented here reveals that SMRs will produce more voluminous and chemically/physically reactive waste than LWRs, which will impact options for the management and disposal of this waste. Although the analysis focuses on only three of dozens of proposed SMR designs, the intrinsically higher neutron leakage associated with SMRs suggests that most designs are inferior to LWRs with respect to the generation, management, and final disposal of key radionuclides in nuclear waste.

Sorption Profile of Low Specific Activity 99Mo on Nanoceria-Based Sorbents for the Development of 99mTc Generators: Kinetics, Equilibrium, and Thermodynamic Studies.

99Mo/99mTc generators play a significant role in supplying 99mTc for diagnostic interventions in nuclear medicine. However, the applicability of using low specific activity (LSA) 99Mo asks for sorbents with high sorption capacity. Herein, this study aims to evaluate the sorption behavior of LSA 99Mo towards several CeO2 nano-sorbents developed in our laboratory. These nanomaterials were prepared by wet chemical precipitation (CP) and hydrothermal (HT) approaches. Then, they were characterized using XRD, BET, FE-SEM, and zeta potential measurements. Additionally, we evaluated the sorption profile of carrier-added (CA) 99Mo onto each material under different experimental parameters. These parameters include pH, initial concentration of molybdate solution, contact time, and temperature. Furthermore, the maximum sorption capacities were evaluated. The results reveal that out of the synthesized CeO2 nanoparticles (NPs) materials, the sorption capacity of HT-1 and CP-2 reach 192 ± 10 and 184 ± 12 mg Mo·g–1, respectively. For both materials, the sorption kinetics and isotherm data agree with the Elovich and Freundlich models, respectively. Moreover, the diffusion study demonstrates that the sorption processes can be described by pore diffusion (for HT-synthesis route 1) and film diffusion (for CP-synthesis route 2). Furthermore, the thermodynamic parameters indicate that the Mo sorption onto both materials is a spontaneous and endothermic process. Consequently, it appears that HT-1 and CP-2 have favorable sorption profiles and high sorption capacities for CA-99Mo. Therefore, they are potential candidates for producing a 99Mo/99mTc radionuclide generator by using LSA 99Mo.

212Pb: Production Approaches and Targeted Therapy Applications.

Over the last decade, targeted alpha therapy has demonstrated its high effectiveness in treating various oncological diseases. Lead-212, with a convenient half-life of 10.64 h, and daughter alpha-emitter short-lived 212Bi (T1/2 = 1 h), provides the possibility for the synthesis and purification of complex radiopharmaceuticals with minimum loss of radioactivity during preparation. As a benefit for clinical implementation, it can be milked from a radionuclide generator in different ways. The main approaches applied for these purposes are considered and described in this review, including chromatographic, solution, and other techniques to isolate 212Pb from its parent radionuclide. Furthermore, molecules used for lead’s binding and radiochemical features of preparation and stability of compounds labeled with 212Pb are discussed. The results of preclinical studies with an estimation of therapeutic and tolerant doses as well as recently initiated clinical trials of targeted radiopharmaceuticals are presented.

Development and Implementation of Nuclear Chemistry Experiments at the Undergraduate Level

There currently exists a significant deficiency in the nuclear chemistry workforce within the United States, despite its importance in areas of medicine, national security, and energy. Poor coverage of this topic across many chemistry departments at the undergraduate level likely contributes to this shortage. Increasing interest and hands-on experience through the availability of laboratory experiments may help alleviate this burden. Herein, we describe a set of nuclear chemistry experiments designed for undergraduate students that can be readily implemented in chemistry departments without specialized programs in this area. These experiments emphasize several concepts related to safe work practices with radioactive materials, such as the effects of distance and shielding. Additionally, the use of a commercially available radionuclide generator was implemented to have students identify an “unknown” radionuclide based on its measured half-life and gamma (γ) spectrum. This laboratory experiment was carried out by two chemistry courses at Cornell University. Student feedback obtained from these sections showed that this hands-on experiment enhanced student understanding of several key concepts of nuclear chemistry and also successfully stimulated interest in this topic. Therefore, this study demonstrates that simple nuclear chemistry laboratory experiments can be implemented in a diverse range of chemistry departments and are effective at fostering student understanding and interest in this topic.

Separation of 44Sc from 44Ti in the Context of A Generator System for Radiopharmaceutical Purposes with the Example of [44Sc]Sc-PSMA-617 and [44Sc]Sc-PSMA-I&T Synthesis.

Today, 44Sc is an attractive radionuclide for molecular imaging with PET. In this work, we evaluated a 44Ti/44Sc radionuclide generator based on TEVA resin as a source of 44Sc. The generator prototype (5 MBq) exhibits high 44Ti retention and stable yield of 44Sc (91 ± 6 %) in 1 mL of eluate (20 bed volumes, eluent—0.1 M oxalic acid/0.2 M HCl) during one year of monitoring (more than 120 elutions). The breakthrough of 44Ti did not exceed 1.5 × 10−5% (average value was 6.5 × 10−6%). Post-processing of the eluate for further use in radiopharmaceutical synthesis was proposed. The post-processing procedure using a combination of Presep® PolyChelate and TK221 resins made it possible to obtain 44Sc-radioconjugates with high labeling yield (≥95%) while using small precursor amounts (5 nmol). The proposed method takes no more than 15 min and provides ≥90% yield relative to the 44Sc activity eluted from the generator. The labeling efficiency was demonstrated on the example of [44Sc]Sc-PSMA-617 and [44Sc]Sc-PSMA-I&T synthesis. Some superiority of PSMA-I&T over PSMA-617 in terms of 44Sc labeling efficiency was demonstrated (likely due to presence of DOTAGA chelator in the precursor structure). It was also shown that microwave heating of the reaction mixture considerably shortened the reaction time and improved radiolabeling yield and reproducibility of [44Sc]Sc-PSMA-617 and [44Sc]Sc-PSMA-I&T synthesis.

Preparation of 191Os–phytate, an in-vivo radionuclide generator, for radiosynovectomy application

191Os is a parent radionuclide with a 15.4 d half-life. It decays by beta emission to 191mIr, which is a radionuclidewith a 4.96s half-life. It decays by the isomeric transition to stable 191Ir, emitting a 129-keV gamma photon. In thisstudy, 191Os–phytate was developed into an in-vivo radionuclide generator for simultaneous radiosynovectomy and imaging. 191Os-hexachloroosmate was used to prepare 191Os–phytate (100 μCi/50 μl) using reaction condition optimization followed by an intraarticular injection to rat knee joints. Also, its distribution and stability were assessed. The imaging of 191Os cation and 191Os–phytate was performed by SPCET. The 191Os–phytate complex was obtained at pH=5.5 with normal saline at room temperature. Radio-TLC showed an overall radiochemical yield of 95-98%. The complex was injected into the rats’ knees, and the whole injected dose remained at the injection site even three days after injection. Due to the stability and retention of the complex in joints approved by biodistribution and imaging studies, the complex is a potential in vivo generator for cavital radiosynovectomy of minor joints.

Improved Sc-44 production in a siphon-style liquid target on a medical cyclotron

In order to use new and promising radiometals for molecular imaging, it is important that they can be obtained as inexpensively and easily as possible. This often requires a cyclotron with solid target hardware or a radionuclide generator, which are not widely available for rarely used radionuclides. Here, we investigate the improved production of 44Sc with a siphon-style liquid target system and compare to our previous work with a simple liquid target. A metal salt solution with a high concentration of natural abundance Ca(NO3)2 (0.14 g/cm3) was irradiated with a medical cyclotron (12 MeV protons; 20 μA). 44Sc was produced via the natCa(p,x)44Sc reaction. As the pressure increase during irradiation was reduced in the siphon-style target, it was possible to irradiate with a higher proton beam current (20 μA) than with the simple liquid target system (7.9 μA). In addition, the saturation yield per μA of 44Sc was increased by a factor of 3.18 ± 0.05 (6.2 ± 0.1 MBq/μA with the siphon target versus 1.94 ± 0.08 MBq/μA with the simple target). This results in an overall increase in 44Sc activity by a factor of 11.

Radiochemical separation of 224Ra from 232U and 228Th sources for 224Ra/212Pb/212Bi generator

The application of diagnostic and therapeutic radionuclides in nuclear medicine has grown significantly and has translated into the increased interest in radionuclide generators and their development. 224Ra and its shorter-lived daughters, 212Pb and 212Bi, are very interesting radionuclides from Targeted Alpha Therapy point of view for treatment of small cancers or metastatic forms. The purpose of the present work was to develop a simple generator for rapid elution of carrier-free 224Ra from 232U or 228Th sources by radiochemical separation based on extraction chromatography with the utilization of a home-made material. The bis(2-ethylhexyl) hydrogen phosphate (HDEHP) extractant was immobilized on polytetrafluroethylene (PTFE) grains and its ability to selectively adsorb 232U and 228Th, with simultaneous high elution recovery of 224Ra, was checked over few years. The 224Ra was quantitatively eluted with small volume (3–5 mL) of 0.1 M HNO3 with low breakthrough (<0.005%) and was used for further milking of 212Bi and 212Pb from DOWEX 50WX12 by 0.75 M and 2.0 M HCl, respectively. The elaborated here methods allowed high recovery of 224Ra, 212Pb and 212Bi radionuclides and their application in radiolabeling of various biomolecules.

Development of a fast characterization setup for radionuclide generators demonstrated by a 227Ac-based generator

Abstract The development of a setup for a fast online characterization of radionuclide generators is reported. A generator utilizing the mother nuclide 227Ac sorbed on a cation exchange resin is continuously eluted by using a peristaltic pump. To allow continuous and pulse-free elution of a large volume over extended time periods a 3D-printed interface designed to remove pressure-oscillations induced by the pump was placed between pump and generator column to ensure undisturbed generator elution. The eluate of the generator is passed through a 3D printed flow cell placed inside a borehole Na(Tl)-scintillation detector for high counting efficiency. Alternatively, a HPGe detector suitable for nuclide identification was used to demonstrate the validity of the online method. The detection system combines conventional gamma-ray spectrometry with fast list mode data acquisition in the Matlab software package. Elution experiments were performed at different flow rates of hydrochloric acid, separating 211Bi (t 1/2 = 2.14 min) free from its parent nuclides. In addition, to prove the versatility of the setup, experiments at different hydrochloric acid concentrations were performed resulting in the elution of pure 211Pb (t 1/2 = 36.1 min) and 223Ra (t 1/2 = 11.43 d), respectively.

Development and validation of a kit formulation of [68Ga]Ga-P15-041 as a bone imaging agent

One of the commonly performed studies in nuclear medicine are bone scans with [99mTc]Tc-methylene diphosphonate (MDP) for detecting various bone lesions, including cancer metastasis. The recent emergence of commercially available 68Ge/68Ga radionuclide generators makes it possible to provide 68Ga-labelled bisphosphonates as positron emission tomography (PET) tracers for bone imaging. Preliminary human studies suggested that [68Ga]Ga-HBED–CC–BP ([68Ga]Ga-P15-041) in conjunction with PET/computed tomography (CT) showed accumulation in known bone lesions, fast clearance from blood and soft tissue, and an ability to provide high contrast images. A simple and efficient lyophilized P15-041 kit formulation for the rapid production of [68Ga]Ga-P15-041 with excellent radiochemical purity (RCP) under ambient temperature without the need for purification is described. It is demonstrated that clinical doses of [68Ga]Ga-P15-041 can be prepared manually within minutes with an excellent purity (> 90%) and readily meet the dose release criteria. When [68Ga]Ga-P15-041 was evaluated in a patient with cancer, the imaging agent clearly showed accumulations in multiple lesions. In conclusion, [68Ga]Ga-P15-041, prepared by a lyophilized kit, might be an excellent bone imaging agent for widespread clinical application.

68 Ga-Labelled Tropane Analogues for the Visualization of the Dopaminergic System.

The development of radiometal‐labelled pharmaceuticals for neuroimaging could offer great potential due to easier handling during labelling and availability through radionuclide generator systems. Nonetheless, to date, no such tracers are available for positron emission tomography, primarily owing to the challenge of crossing the blood–brain barrier (BBB) and loss of affinity through chelator attachment. We have prepared a variety of 68Ga‐labelled phenyltropanes showing that, through a simple hydrocarbon‐linker, it is possible to introduce a chelator onto the lead structure while maintaining its high affinity for hDAT (human dopamine transporter) and simultaneously achieving adequate lipophilicity. One of the candidates, [68Ga]Ga‐HBED‐hexadiyne‐tropane, showed an IC50 value of 66 nM, together with a log D 7.4 of 0.96. A μPET study in a hemi‐parkinsonian rat model showed a fast wash‐out of the tracer, and no specific uptake in the brain, thus implying an inability to penetrate the BBB.

Examination of metallic impurities of 68Ge/68Ga generators used for radioactive labeling of peptides in clinical PET applications.

Especially suitable for PET due to its nuclear physical and radiochemical properties, the positron emitter Gallium-68 (Ga) occurs by electron capture from Germanium-68 (Ge). In such a radionuclide generator, the germanium is bound to an insoluble, inert column matrix and forms a secular radioactive balance with 68Ga obtained in the hour. As a result of the limited radiochemical selectivity of the elution process, the eluate obtained is basically contaminated with the main nuclide traces, so that the eluate becomes a mixture of Ga and Ge radionuclides. Also, the generator eluate contains a number to metal cations that reduce specific radioactivity and can compete with 68Ga. The presence of toxic metal that can be found in the eluate carries the risks of contamination at every step from the production of generators to radiopharmaceutical production. In our study, by collecting the eluate of the Ge/Ga generators used with different identities in different centers in Turkey, we report comparative analysis of metal contamination in the generator eluate. The eluates of 68Ge/68Ga generators to five different identities were collected. Eluates were analyzed by inductively coupled plasma-mass spectrometry. As a result, each generator contains metallic impurities different from its certificate.

A reverse 230U/226Th radionuclide generator for targeted alpha therapy applications

PurposeThorium-226 (half-life 30.6 m) is a radionuclide of interest for use in targeted alpha therapy applications. Due to its short half-life, 226Th must be provided through a radionuclide generator system from its parent 230U (20.8 d). Furthermore, as the half-life of 226Th is very short, it should be provided in a form that is directly amenable to use in biomedical applications. MethodsA reverse radionuclide generator system was developed employing a DGA extraction chromatography column. A 230U/226Th parent/daughter solution in equilibrium is added to a DGA column in >6 M HCl. The parent 230U is eluted first in 0.1 M HNO3 followed by elution of 226Th in 0.1 M citrate buffer pH 5. ResultsThorium-226 was recovered from the radionuclide generator column with >96% yield. Greater than 99.5% of the 230U parent was isolated for reuse in the generator. Long term evaluation over six weeks demonstrated consistent supply of 226Th with greater than 99.5% radionuclidic purity. The only contaminant found in the final product was 230U (<0.5%). ConclusionsThe reverse radionuclide generator described herein was shown to be a feasible method for providing 226Th in high yield, purity and in a chemical form that is amenable for direct use in biomedical applications.

Preparation and Preliminary Evaluation of 68Ga-Acridine: An Attempt to Study the Potential of Radiolabeled DNA Intercalator as a PET Radiotracer for Tumor Imaging.

Acridine is a well-known DNA intercalator and thereby gets easily inserted within DNA. As uncontrolled rapid cell division is one of the primary characteristics of the tumors, it is expected that acridine or its suitable derivatives will have preferential accumulation in the tumorous lesions. Therefore, an attempt was made to radiolabel an acridine derivative with 68Ga and study the potential of the 68Ga-acridine complex as a PET agent for tumor imaging. 9-aminoacridine was coupled with p-NCS-benzyl-DOTA to render it suitable for labeling with 68Ga. The purified acridine-DOTA conjugate was radiolabeled with 68Ga, eluted from a 68Ge/68Ga radionuclide generator. Various radiolabeling parameters were optimized and the stability of the radiolabeled preparation was studied. The biological behavior of the 68Ga-acridine complex was studied both in vitro and in vivo using Raji cell line and fibrosarcoma tumor bearing Swiss mice, respectively. 68Ga-acridine complex was obtained with ~100% radiochemical purity under the optimized reaction conditions involving incubation of 2mg/mL of ligand at 100°C for 30 minutes. The complex maintained a radiochemical purity of >95% in normal saline and >65% in human blood serum at 3h post-incubation. In vitro cellular study showed (3.2±0.1)% uptake of the radiotracer in the Raji cells. Biodistribution study revealed significant tumor accumulation [(11.41±0.41)% injected activity in per gram] of the radiotracer within 1h postadministration along with uptake in other non-target organs such as, blood, liver, GIT kidney etc. Conclusion: The present study indicates the potential of 68Ga-acridine as a PET agent for imaging of tumorous lesions. However, further detailed evaluation of the agent is warranted to explore its actual potential.

Development and long-term evaluation of a new 68Ge/68Ga generator based on nano-SnO2 for PET imaging

Radionuclide generator systems can routinely provide radionuclides on demand such as 68Ga produced by a 68Ge/68Ga generator without the availability of an on-site accelerator or a research reactor. Thus, in this work nano-SnO2 was used to develop a new 68Ge/68Ga generator which was evaluated over a period of 17 months and 305 elution cycles. The elution yield was 91.1 ± 1.8% in the first 7 mL (1 M HCl as eluent) when the generator was new and then it decreased with time and use to 73.8 ± 1.9%. Around 80% of the elutable 68Ga activity was obtained in 1 mL and the 68Ge content in the eluate did not exceed 1 × 10–4% over the investigation period when it was eluted regularly. The described generator provided adequate results for radiolabelling of DOTA-TOC with direct use of eluate. In addition, [68Ga]Ga-DOTA-TOC was tested satisfactorily for in vivo tumor detection by microPET/CT imaging in a lung cancer mouse model.

Modelling of the 177mLu/177Lu radionuclide generator

In order to determine the potential of 177mLu/177Lu radionuclide generator in 177Lu production it is important to establish the technical needs that can lead to a clinically acceptable 177Lu product quality. In this work, a model that includes all the processes and the parameters affecting the performance of the 177mLu/177Lu radionuclide generator has been developed. The model has been based on the use of a ligand to complex 177mLu ions, followed by the separation of the freed 177Lu ions. The dissociation kinetics of the Lu-ligand complex has been found to be the most crucial aspect governing the specific activity and 177mLu content of the produced 177Lu. The dissociation rate constants lower than 1*10-11 s-1 would be required to lead to onsite 177Lu production with specific activity close to theoretical maximum of 4.1 TBq 177Lu/mg Lu and with 177mLu content of less than 0.01%. Lastly, the calculations suggest that more than one patient dose per week can be supplied for a period of up to 7 months on starting with the 177mLu produced using 3 g Lu2O3 target with 60% 176Lu enrichment. The requirements of the starting 177mLu activity production needs to be adapted depending on the required patient doses, and the technical specifications of the involved 177mLu-177Lu separation process.

Solid phase extraction-based separation of the nuclear isomers 177mLu and 177Lu

A solid phase extraction based 177mLu-177Lu separation method has been investigated for its feasibility to be used in the radionuclide generator. The use of 2,2′,2”-(10-(2,6-dioxotetrahydro-2H-pyran-3-yl)-1,4,7,10-tetraazacyclododecane-1,4,7-triyl)triacetic acid, (DOTAGA-anhydride) allowed grafting of DOTA (1,4,7,10-tetraazacyclododecane N,N′,N″,N‴-tetraacetic acid) complex on the surface of commercially available amino propyl silica. The grafting of DOTA has been confirmed by several characterization techniques. The thermogravimetric analysis reveals that the 0.33 mmol DOTA groups have been grafted per gram of silica. However, during the Lu ion complexation, a 10 times lower Lu adsorption capacity of 0.03 mmol g−1 could be achieved under the studied reaction conditions. The results indicate that the grafting of DOTA on solid affects the Lu coordination and also influences the kinetics of Lu-DOTA complexation. The weak coordination resulted in high 177mLu leakage, while the unreacted DOTA groups interfer with the 177Lu release. This is evident from the 0.3% 177mLu leakage combined with a177Lu extraction efficiency of 25%. Overall, the results show a177mLu-177Lu separation with a maximum 177Lu/177mLu activity ratio of 25. But this is still far away from clinically acceptable activity ratio of 10,000 for which future work is recommended.

Construction of a thorium/actinium generator at the Canadian Nuclear Laboratories

Targeted Alpha Therapy (TAT) has demonstrated considerable promise in the treatment of a range of cancers in both preclinical and, more recently clinical research. In particular, work with the alpha-emitting radionuclide 225Ac has been effectively employed due to the relatively rapid decay cascade that leads to 4 alpha and 2 beta emissions. One limitation for TAT has been caused by access to the vital radionuclide. Traditionally, 225Ac has been sourced from thorium/actinium generators based on the alpha decay of stockpiles of 229Th. 229Th is itself the alpha-decay product from 233U. Due to proliferation issues associated with 233U, only three thorium/actinium generators have been reported in the literature, capable of supporting clinical research. This paper describes the construction and operation of a thorium/actinium radionuclide generator at the Canadian Nuclear Laboratories, capable of supporting preclinical and limited clinical research in the area of TAT. Thorium was recovered and purified by a combination of anion exchange and extraction chromatography from aged 233U stockpiles. A separation scheme for 225Ra and 225Ac has been developed, based upon the chemical composition of the thorium material to allow for regular, routine milkings capable of supplying up to 3.7 GBq (100 mCi) of radiochemically pure 225Ac annually. This routine separation is accomplished using a combination of anion exchange chromatography to separate Ac and Ra isotopes from Th and extraction chromatography employing TEVA and DGA-N resins to separate actinium from radium and breakthrough thorium.