Radionuclide Generator Research Articles

Environmental radioactivity aspects of recent nuclear accidents associated with undeclared nuclear activities and suggestion for new monitoring strategies

Recently environmental radionuclide signals were observed in the atmosphere which could be associated with undeclared nuclear activities, not directly connected with development of nuclear weapons. Large-scale contamination of European air with Ru-106 observed in 2017 may represent such an accident, which was probably associated with the Mayak nuclear fuel reprocessing facility in the Chelyabinsk region of Russia. A recently announced nuclear accident at Nyonoska in the Archangelsk region may represent an undeclared nuclear activity associated with the development of a nuclear jet engine which could be based on radionuclide energy generator or a small nuclear reactor. It is concluded that environmental radioactivity impacts associated with recent nuclear activities create new challenges for fast and reliable national and international monitoring systems which would require development of new monitoring strategies.

Harvesting62Zn from an aqueous cocktail at the NSCL

The radionuclide62Zn was obtained by “isotope harvesting” and separated from other co-produced species. The principle of a medical radionuclide generator was demonstrated by isolating pure62Cu, which is generated by the decay of its parent62Zn.

Radionuclide generator-based production of therapeutic\xa0177Lu from its long-lived isomer 177mLu

BackgroundIn this work, a lutetium-177 (177Lu) production method based on the separation of nuclear isomers, 177mLu & 177Lu, is reported. The 177mLu-177Lu separation is performed by combining the use of DOTA & DOTA-labelled peptide (DOTATATE) and liquid-liquid extraction.MethodsThe 177mLu cations were complexed with DOTA & DOTATATE and kept at 77 K for periods of time to allow 177Lu production. The freed 177Lu ions produced via internal conversion of 177mLu were then extracted in dihexyl ether using 0.01 M di-(2-ethylhexyl) phosphoric acid (DEHPA) at room temperature. The liquid-liquid extractions were performed periodically for a period up to 35 days.ResultsA maximum 177Lu/177mLu activity ratio of 3500 ± 500 was achieved with [177mLu]Lu-DOTA complex, in comparison to 177Lu/177mLu activity ratios of 1086 ± 40 realized using [177mLu]Lu-DOTATATE complex. The 177Lu-177mLu separation was found to be affected by the molar ratio of lutetium and DOTA. A 177Lu/177mLu activity ratio up to 3500 ± 500 was achieved with excess DOTA in comparison to 177Lu/177mLu activity ratio 1500 ± 600 obtained when lutetium and DOTA were present in molar ratio of 1:1. Further, the 177Lu ion extraction efficiency, decreases from 95 ± 4% to 58 ± 2% in the presence of excess DOTA.ConclusionThe reported method resulted in a 177Lu/ 177mLu activity ratio up to 3500 after the separation. This ratio is close to the lower end of 177Lu/177mLu activity ratios, attained currently during the direct route 177Lu production for clinical applications (i.e. 4000–10,000). This study forms the basis for further extending the liquid-liquid extraction based 177mLu-177Lu separation in order to lead to a commercial 177mLu/177Lu radionuclide generator.

Radiolysis reduction in liquid solution targets for the production of 89Zr

Increased interest in radiometals for nuclear medicine and imaging can be hampered by radionuclide supply. 89Zr for example, is a PET imaging nuclide for which no radionuclide generator exists. One method to produce 89Zr involves irradiating aqueous solutions of yttrium nitrate salt on small medical cyclotrons. However, in irradiating these solutions the radiolysis of water can cause significant H2 and O2 gas buildup, which can eventually rupture a sealed target vessel.We examine the role of nitrate and nitrite in radiolysis. Here, we find that using copper-coated cadmium pellets to chemically reduce nitrate to nitrite in solution prior to irradiation can reduce in-target radiolysis by approximately 60% as compared to other published methods of radiolysis reduction, but only in acidic solutions. We hypothesize that during irradiation, nitrate is converted to nitrite, consuming free radicals which would otherwise be available to eliminate molecular gas species. Performing this conversion before irradiation may limit the consumption of these beneficial free radicals.

Solid state synthesis of mesoporous alumina: A viable strategy for preparation of an advanced nanosorbent for 99Mo/99mTc generator technology

Abstract Solid state reactions are becoming increasingly relevant as a technique for clean and efficient synthesis of advanced materials of immense technological relevance. Adsorption based column chromatographic processes are among of the most viable routes for radiochemical separation of medically important radioisotopes for clinical benefits. The fundamental and great challenge in this regard lies in development of highly efficient sorbent materials. Owing to their intrinsically high surface area and large pore volume, ordered mesoporous metal oxides may be the answer in the future for column chromatographic radiochemical separations. This work reports the synthesis of mesoporous alumina with greatly enhanced textural characteristics through solid state reaction of aluminium isopropoxide with glucose, followed by high temperature calcination. The nanostructured material (2–4 nm crystallite size) could be synthesized with large surface area (287 ± 16 m 2 /g) and narrow pore size distribution with mean pore radius of 1.8 ± 0.3 nm. When evaluated as Mo adsorbent for use in preparation of 99 Mo/ 99m Tc generator, mesoporous alumina displayed >10 times higher sorption capacity compared to conventionally used bulk alumina. A clinical-scale 99 Mo/ 99m Tc generator (20.4 GBq) was developed using this sorbent material. Technetium-99 m could be regularly eluted from this generator over a prolonged period of 2 weeks with >80% elution yield and it met all the requirements for clinical use. Overall, this study is expected provide inspiration to explore the relatively new terrain of solid state synthesis for large-scale preparation of ‘next generation’ sorbent materials for widespread utilization in radionuclide generator technology.

Safety and efficacy of targeted alpha therapy with 213Bi-DOTA-substance P in recurrent glioblastoma.

Twenty patients with recurrent GBM were included into the study following a standard therapy. 1-2 intracavitary or intratumoral port-a-cath systems were stereotactically inserted. Patients were treated with 1-7 doses of 213Bi-DOTA-Substance P (213Bi-DOTA-SP) in 2-month intervals. 68Ga-DOTA-Substance P (68Ga-DOTA-SP) was co-injected with 213Bi-DOTA-SP to assess the biodistribution using PET/CT. Therapeutic response was monitored with performance status and MRI imaging. Treatment with activity up to 11.2GBq 213Bi-DOTA-SP was well tolerated with only mild and transient adverse reactions. The median progression free survival was 2.7months. The median overall survival from the first diagnosis was 23.6months and median survival after recurrence was 10.9months. The median survival time from the start of 213Bi-DOTA-SP was 7.5months. Treatment of recurrent GBM with 213Bi-DOTA-SP is safe and well tolerated. The median overall survival after recurrence of 10.9months compares favorably to the available alternative treatment options. Once the supply of high activity 225Ac/213Bi radionuclide generators is secured, targeted alpha therapy with 213Bi-DOTA-SP may evolve as a promising novel option to treat recurrent GBM.

68Ga-DOTA-TATE—a source of eye lens exposure for nuclear medicine department workers

Introduction. Obtaining 68Ga-DOTA-TATE (a radioconjugate consisting of the somatostatin analogue tyrosine-3-octreotate (Tyr3-octreotate or TATE) labelled with the positron emission tomography tracer gallium 68Ga via the macrocyclic chelating agent dodecanetetraacetic acid (DOTA)) is a complex process and, as with any radiopharmaceutical whose basis is a short-lived radionuclide generator, it is based on a sequence of procedures beginning from the 68Ge/68Ga generator elution, labelling ligands with a radioisotope, dispensing doses of 68Ga-DOTA-TATE for patients and finally injection of the preparation to patients. The complexity of this process may contribute to an increased exposure of eye lenses of the staff who perform the above-mentioned procedures, which is especially important at a time when the dose limit on the lens of the eye is being reduced from 150 to 20 mSv yr–1. Objective. The work presents the exposure of eye lenses of the personnel of a nuclear medicine department who prepare and inject 68Ga-DOTA-TATE. Materials and methods. Radiochemists and nurses were monitored by dosimetry measurements with thermoluminescent detectors (TLDs). Results. The values of Hp(3)/A—normalised personal eye dose equivalent recorded in the group of radiochemists during the procedure of dispensing the doses of 68Ga-DOTA-TATE for patients exceeded the value of 274 μSv/GBq. Conclusions. The estimated annual Hp(3) values may exceed 20 mSv, which is particularly important due to the fact that procedures using the 68Ga radioactivity are only a small part of the daily professional activity of the staff, resulting from the performance of other procedures that require the use of other radioisotopes.

Physical and chemical processes on the 212Pb radionuclide production for nuclear medicine

An automated 212Pb radionuclide generator with 228Th as a parent isotope has been designed. The generator operation principle is based on transport by an airflow of gaseous 220Rn radionuclide emanating from DOWEX ion exchange resin containing 228Th into a helix-shaped collector vessel where post-decay 212Pb is collected and deposited on the collector walls. After a 72-hour operation cycle of the generator, sampling of 212Pb in the form of solution in 0.1M HCl is executed with approximately 30% yield of 212Pb. The generator is remotely controlled by a PC-based automatic system. The generator is reloaded once in a few years. The generator is intended for medicobiological and radiochemical investigations in the field of developing antineoplastic drugs of targeted delivery.

Получение позитрон-излучающего радионуклида 68Ga: радиохимическая схема радионуклидного генератора 68Ge → 68Ga

68Ga (T1 / 2 = 68 min) in complexes with peptides is used in positron emission tomography for diagnostics of neuroendocrine tumors. The most promising strategy for 68Ga production is usage of the radionuclide generator 68Ge → 68Ga. In this research, the sorption behavior of Ge(IV) and Ga (III) has been studied. The distribution coefficients (Kd) of Ge(IV) on the anion exchange (Dowex 1×8) and cation exchange (Dowex 50×8) resins in various ethanedioic and hydrochloric acid solutions were determined. For each ion exchange resin, four series of measurements were carried out, in which the concentration of oxalic acid was fixed (0.001 M, 0.003 M, 0.005 M, 0.01 M), and the concentrations of hydrochloric acid ranged from 0 to 3 M. Based on the distribution coefficients, the chemical scheme of the radionuclide generator 68Ge → 68Ga has been developed. The chemical system is based on the anion exchange resin Dowex 1×8 and mixture of 0.005 M C2H2O4 / 0.33 M HCl. Several types of the generators with direct and reverse mode of elution were tested and the optimal scheme was determined. Elution of the generators was performed once a day with 8 ml of 0.005 M C2H2O4 / 0.33 M HCl solution. The 68Ga yield and the 68Ge breakthrough are comparable for all the systems.

Quality assurance of Mo-99/Tc-99m radionuclide generators

Abstract Gamma-ray spectrometry analyses of the radionuclide content of eluate from two Mo-99/Tc-99m radionuclide generators POLTECHNET have been performed. The relative activities of 99Mo 103Ru and 131I radioisotopes with respect to the activity of 99mTc at different time intervals after the primary pertechnetate elution of the generators have been analyzed. The relative activities of the isotopes were determined and compared to the radionuclidic purity requirements for 99mTc.

Evaluation of Standardized Uptake Value applying Prompt Gamma Correction on 68Ga-DOTATOC PET/CT Image

68Ga was eluted from a 68Ge/68Ga radionuclide generator. 68Ga decays into 68Zn, with a half life=67.8min. The decay is 88.9 % by β+ and 11.1 % by EC. The main β+ decay (87.7 %) is to the ground level of 68Zn and it is a pure positron emission branch. A small fraction decays β+ (1.2 %) into an excited level of 68Zn, which promptly decays into the ground level with a γ (1.077 Mev). This can constitute prompt gamma contamination in the PET data, if the 1.077 Mev γ has a scatter interaction in the patient, and generates a lower energy γ in coincidence with the positron annihilation pair. The purpose of this study was to evaluate standardized uptake value(SUV) before and after applying prompt gamma rays correction on 68Ga-DOTATOC PET/CT image. Fifty patient underwent PET/CT 1 hour after injection of the 68Ga-DOTATOC. The SUVmax and SUVmean of lesions and normal tissues (Pituitary, Lung, Liver, Spleen, Kidney, Intestine) were evaluated before and after applying prompt gamma correction on 68Ga-DOTATOC PET/CT image. Additionally, the SUVmax of each lesions and SUVmean of the soft tissues were measured on images. and target to background ratios (TBR) were calculated as quantitative indices. Among 15 patients, 25 of lesions (Pancreas, Liver, Thoracic Spine, Brain) with increased uptake on 68Ga-DOTATOC PET/CT image. SUVmax and SUVmean were increased in lesion site and normal tissue after prompt gamma rays correction. TBR was 51.51 ± 49.28 and 55.50 ± 53.12 before and after prompt gamma rays correction, respectively. (p<0.0001)

Utilization of (p, 4n) reaction for 86Zr production with medium energy protons and development of a 86Zr → 86Y radionuclide generator

Utilization of (p, 4n) reaction channel for the production of medical radionuclides became very attractive with commercial availability of medium energy cyclotrons. Significantly higher yields and radionuclidic purity may open new perspectives for several novel and some of the radionuclides previously have not been considered due to production difficulties. In present work, we show the proof-of-principle study on the production of 86Y for Positron Emission Tomography imaging via radionuclide generator 86Zr → 86Y. Production suitability of 86Zr from natural yttrium target and radiochemical separation strategies were tested. In addition, two generator systems were proposed and evaluated.

The potential of radiolabeled chemotherapeutics in tumor diagnosis: Preliminary investigations with 68 Ga-gemcitabine.

Preclinical Research & Development Gemcitabine, a nucleoside analog, is a well-known chemotherapeutic drug that is used either alone or with other agents to treat a wide variety of cancers. The aim of the present work was to evaluate the potential of 68 Ga-labeled gemcitabine for its application in positron emission tomography (PET) imaging of tumorous lesions. Gemcitabine was coupled with p-NCS-benzyl-DOTA in order to facilitate radiolabeling with 68 Ga. The gemcitabine-p-NCS-benzyl-DOTA was radiolabeled with 68 Ga, obtained from a 68 Ge/68 Ga radionuclide generator. The radiolabeled product was characterized by high performance liquid chromatography (HPLC) and its tumor specificity was evaluated by biodistribution studies in Swiss mice bearing fibrosarcoma tumors. Preliminary bioevaluation study showed good tumor uptake within 1 hr post-administration [2.5% Injected Activity (IA) per g of tumor] with rapid renal clearance (>90% IA) and a high tumor to muscle ratio. 68 Ga-gemcitabine may have potential as a PET agent for tumor imaging.

Antitumor effects of radionuclide treatment using \u03b1-emitting meta-211At-astato-benzylguanidine in a PC12 pheochromocytoma model

PurposeTherapeutic options for patients with malignant pheochromocytoma are currently limited, and therefore new treatment approaches are being sought. Targeted radionuclide therapy provides tumor-specific systemic treatments. The β-emitting radiopharmaceutical meta-131I-iodo-benzylguanidine (131I-MIBG) provides limited survival benefits and has adverse effects. A new generation of radionuclides for therapy using α-particles including meta-211At-astato-benzylguanidine (211At-MABG) are expected to have strong therapeutic effects with minimal side effects. However, this possibility has not been evaluated in an animal model of pheochromocytoma. We aimed to evaluate the therapeutic effects of the α-emitter 211At-MABG in a pheochromocytoma model.MethodsWe evaluated tumor volume-reducing effects of 211At-MABG using rat pheochromocytoma cell line PC12 tumor-bearing mice. PC12 tumor-bearing mice received intravenous injections of 211At-MABG (0.28, 0.56, 1.11, 1.85, 3.70 and 5.55 MBq; five mice per group). Tumor volumes were evaluated for 8 weeks after 211At-MABG administration. The control group of ten mice received phosphate-buffered saline.ResultsThe 211At-MABG-treated mice showed significantly lower relative tumor growth during the first 38 days than the control mice. The relative tumor volumes on day 21 were 509.2% ± 169.1% in the control mice and 9.6% ± 5.5% in the mice receiving 0.56 MBq (p < 0.01). In addition, the mice treated with 0.28, 0.56 and 1.11 MBq of 211At-MABG showed only a temporary weight reduction, with recovery in weight by day 10.Conclusion211At-MABG exhibited a strong tumor volume-reducing effect in a mouse model of pheochromocytoma without weight reduction. Therefore, 211At-MABG might be an effective therapeutic agent for the treatment of malignant pheochromocytoma.

An evaluation of activation and radiation damage effects for the European Spallation Source Target

ABSTRACT The details concerning the generation and accumulation of radionuclides produced by spallation reactions within the neutron generation target to be installed at the European Spallation Source are described. The resulting radiation damage for the target tungsten blocks constituents as well as the target encasing is evaluated and explicit results are given for the effects of knock-on atoms as well as those resulting from the accumulation of light and volatile hydrogen and helium isotopes resulting from spallation reactions. Their effects on the mechanical properties are discussed on the light of available experimental data and, on this basis, an estimation of the target lifetime is given.

Sorption behaviour of pertechnetate on oxidized and reduced surface of activated carbon

Objectives of this research are preparation and characterization of oxidised and reduced activated carbon, and investigation of pertechnetate sorption on these activated carbons. Activated carbon could be an alternative interface in production of radionuclide generator with high Mo concentration, such as it is case for cyclotron or photoreaction producing 99mTc from 100Mo target. Sorption processes are investigated through radioisotope indication and strongly depend on pH and presence of competitive anions of the environment. Oxidative treatment leads to an increase in pertechnetate adsorption. Desorption of pertechnetate from the activated carbon surface is almost quantitative.

A study of separation and solidification of group II nuclides in waste salt delivered from the pyrochemical process of used nuclear fuel

If group II nuclides, which contain high heat-generative elements, in waste salt are fabricated into a waste form rich in group II nuclides, the waste form can be used in radionuclide thermoelectric generator applications. For this reason, the separation of group II nuclides in salt (LiCl, LiCl-KCl) was conducted, after which a waste form rich in them was fabricated. In this study, group II nuclide chlorides in salt were effectively separated into a carbonate or oxychloride form, and the separated nuclides were successfully fabricated into a homogenous and stable glass waste form with high contents (45–50 wt%) of these nuclides.

Separation of nuclear isomers for cancer therapeutic radionuclides based on nuclear decay after-effects

177Lu has sprung as a promising radionuclide for targeted therapy. The low soft tissue penetration of its β− emission results in very efficient energy deposition in small-size tumours. Because of this, 177Lu is used in the treatment of neuroendocrine tumours and is also clinically approved for prostate cancer therapy. In this work, we report a separation method that achieves the challenging separation of the physically and chemically identical nuclear isomers, 177mLu and 177Lu. The separation method combines the nuclear after-effects of the nuclear decay, the use of a very stable chemical complex and a chromatographic separation. Based on this separation concept, a new type of radionuclide generator has been devised, in which the parent and the daughter radionuclides are the same elements. The 177mLu/177Lu radionuclide generator provides a new production route for the therapeutic radionuclide 177Lu and can bring significant growth in the research and development of 177Lu based pharmaceuticals.

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant, as well as the activation of coolant and impurities, were analysed for a fluoride salt-cooled high-temperature reactor (FHR) system, and the activity of radionuclides accumulated in the coolant during normal operation was calculated. The release rate (release fraction per unit time) of fission products was calculated with STACY code, which is modelled mainly based on the Fick’s law, while the activation of coolant and impurities was calculated with SCALE code. The accumulation of radionuclides in the coolant has been calculated with a simplified model, which is generally a time integration considering the generation and decay of radionuclides. The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown. In operation condition, health-impacts related nuclides such as 3H, and 14C originate from the activation of lithium and coolant impurities including carbon, nitrogen, and oxygen. According to the calculated effective cross sections of neutron activation, 6Li and 14N are the dominant 3H production source and 14C production source, respectively. Considering the high production rate, 3H and 14C should be treated before being released to the environment.

Clinical Translation and First In-Human Use of [44Sc]Sc-PSMA-617 for PET Imaging of Metastasized Castrate-Resistant Prostate Cancer.

Background: Various trivalent radiometals are well suited for labeling of DOTA-conjugated variants of Glu-ureido-based prostate-specific membrane antigen (PSMA) inhibitors. The DOTA-conjugate PSMA-617 has proven high potential in PSMA radioligand therapy (PSMA-RLT) of prostate cancer as well as PET imaging when labeled with lutetium-177 and gallium-68 respectively. Considering the relatively short physical half-life of gallium-68 this positron emitter precludes prolonged acquisition periods, as required for pre-therapeutic dosimetry or intraoperative applications. In this context, the positron emitter scandium-44 is an attractive alternative for PET imaging. We report the synthesis of [44Sc]Sc-PSMA-617 as radiopharmaceutical with generator produced scandium-44, its in vitro characterization and clinical translation as part of a first in-human study.Methods: Scandium-44 was obtained from a 44Ti/44Sc radionuclide generator. PSMA-617 was labeled with 142.4±12.7 MBq of scandium-44 in analogy to [68Ga]Ga-PSMA-617 and evaluated in vitro and in cell studies using PSMA+ LNCaP cells. A first-in-human investigation was subsequently carried out in a cohort of 4 patients (mean age 70±1.8 a) registered for [177Lu]Lu-PSMA-617 therapy. 50.5±9.3 MBq (40 µg, 38.4 nmol) [44Sc]Sc-PSMA-617 were applied via intravenous injection (i.v.), respectively. A Siemens Biograph 2 PET/CT system was used to acquire initial dynamic PET data (30 min) of abdomen in list mode followed by static PET/CT data (skull to mid-thigh) at 45 min, 2 and 18 h post-injection (p.i.). For quantitative analysis, dynamic images were reconstructed as 6 data sets of 300 s each. The noise ratio was measured in liver, lung and an additional region outside the body. SUV values in different organs and lesions were measured and compared to [68Ga]Ga-PSMA-11 data of the same patients. Residence times and organ absorbed doses were calculated using OLINDA/EXM software.Results: Quantitative radiochemical yields of ≥98 % were achieved using 18 nmol of PSMA-617 after 20 min at 95 °C with apparent molar activity of 6.69±0.78 MBq/nmol. Following purification, >99 % radiochemical purity was obtained. [44Sc]Sc-PSMA-617 showed high stability (>95 %) in serum for 24 h. The binding affinity and internalization fraction were determined in PSMA+ LNCaP cells (IC50 = 4.72±0.7 nM and internalization fraction: 15.78±2.14 % IA/106 LNCaP cells) and compared to [68Ga]Ga-PSMA-11 (12.0±2.8 nM and 9.47±2.56 % IA/106 LNCaP cells). Physiological tracer uptake was observed in kidneys, liver, spleen, small intestine, urinary bladder, and salivary glands and pathological uptake in both soft and skeletal metastases. SUV values were significantly lower in the kidneys (14.0) compared to [68Ga]Ga-PSMA-11 OET (30.5). All other measured SUV values did not show a statistically significant difference. Tumor to liver ratios were found to lie between 1.9 and 8.3 for [68Ga]Ga-PSMA-11 and between 2.5 and 8.8 for [44Sc]Sc-PSMA-617 after 120 min. For [44Sc]Sc-PSMA-617 the ratios were higher and no statistically significant differences were observed. Total and % activity were highest in liver followed by kidneys, spleen, small intestine and salivary glands. Rapid wash out was seen in liver and spleen and gradually over time in kidneys. Kidneys received the highest radiation absorbed dose of 0.354 (0.180-0.488) mSv/MBq. No adverse pharmacological effects were observed.Conclusion: In conclusion [44Sc]Sc-PSMA-617 PET is suitable for PET imaging of prostate cancer tissue. [44Sc]Sc-PSMA-617 shows promise to enable pre-therapeutic dosimetry in clinical settings. However, the clinical advantages for individual dosimetry or other applications like intraoperative applications have to be investigated in further studies.