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

BACKGROUND: PET/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 FINDINGS: Reliable 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. CONCLUSION: The described synthesis method enables reproducible, automated in-house production of [68Ga]Ga-DOTA-exendin-4 for routine clinical application.

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.

Current State of 44Ti/44Sc Radionuclide Generator Systems and Separation Chemistry.

In recent years, there has been an increased interest in 44Ti/44Sc generators as an onsite source of 44Sc for medical applications without needing a proximal cyclotron. The relatively short half-life (3.97 hours) and high positron branching ratio (94.3%) of 44Sc make it a viable candidate for positron emission tomography (PET) imaging. This review discusses current 44Ti/44Sc generator designs, focusing on their chemistry, drawbacks, post-elution processing, and relevant preclinical studies of the 44Sc for potential PET radiopharmaceuticals.

[Long-term Quality Control Test of a 68Ge/68Ga Generator].

Quality of the 68Ga solution eluted from a 68Ge/68Ga generator (Galli Eo®) was evaluated. Elution was performed 488 times from 21 to 484 days (ca. 15 months) after the calibration date. The eluted 68Ga activity was 329 MBq to 1,148 MBq, which decreased with the decay of 68Ge, but no significant change was observed in the elution yields (62.8±2.2%). The half-life of the eluted radioactivity was 67.8±0.1 min, the radionuclide purity of 68Ga was ≥99.9%, the 68Ge breakthrough was 0.000024 ±0.000004%, and the radiochemical purity of 68Ga3+ was 99.7±0.2%. Fe and Zn were detected as metal impurities in the eluent, but both were ≤10 µg/GBq. The endotoxin concentration of the eluate was ≤5 EU/mL, and the eluent passed the sterility test. These results show that the generator can stably provide 68Ga solution over a 15-month period.

Calculations for the nuclear reaction cross-sections via α-particle induced reactions on 65Cu to produce impurity free 67Ga for medical applications.

Diagnostics field is facilitated with advancements enacted in anatomic imaging (cross-sectional modalities). Radionuclide scans (imaging) escorted by 67Ga are extensively beneficial in bone scintigraphy and recognition of prosthetic joint failure. Present work comprises the data concerning 67Ga production via α-particle induced nuclear reactions, TTY (thick target yield) and impurity analysis. Experimental measurements regarding 67Ga production are analyzed through a comparative study performed with calculations of theoretical model codes (TALYS-1.95, EMPIRE-3.2.3 and ALICE-IPPE). A data set of recommended cross-sections was generated and utilized to deduce TTY. The contribution of radionuclidic impurities is canvassed to suggest an energy region to produce impurity free 67Ga for medical applications.

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

BACKGROUND/AIM: 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. MATERIALS AND METHOD: 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. RESULTS AND CONCLUSION: As a result, each generator contains metallic impurities different from its certificate.

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

PURPOSE: Thorium-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. METHODS: A 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. RESULTS: Thorium-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%). CONCLUSIONS: The 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.

Rubidium-82 generator yield and efficiency for PET perfusion imaging: Comparison of two clinical systems.

INTRODUCTION: Strontium-82/Rubidium-82 (82Sr/82Rb) generators are used widely for positron emission tomography (PET) imaging of myocardial perfusion. In this study, the 82Rb isotope yield and production efficiency of two FDA-approved 82Sr/82Rb generators were compared. METHODS: N = 515 sequential daily quality assurance (QA) reports from 9 CardioGen-82® and 9 RUBY-FILL® generators were reviewed over a period of 2 years. A series of test elutions was performed at different flow-rates on the RUBY-FILL® system to determine an empirical correction-factor used to convert CardioGen-82® daily QA values of 82Rb activity (dose-calibrator 'maximum' of 50 mL elution at 50 mL·min-1) to RUBY-FILL® equivalent values (integrated 'total' of 35 mL elution at 20 mL·min-1). The generator yield (82Rb) and production efficiency (82Rb yield/82Sr parent activity) were measured and compared after this conversion to a common scale. RESULTS: At the start of clinical use, the system reported 82Rb activity from daily QA was lower for CardioGen-82® vs RUBY-FILL® (2.3 ± 0.2 vs 3.0 ± 0.2 GBq, P < 0.001) despite having similar 82Sr activity. Dose-calibrator 'maximum' (CardioGen-82®) values were found to under-estimate the integrated 'total' (RUBY-FILL®) activity by ~ 24% at 50 mL·min-1. When these data were used to convert the CardioGen-82 values to a common measurement scale (integrated total activity) the CardioGen-82® efficiency remained slightly lower than the RUBY-FILL® system on average (88 ± 4% vs 95 ± 4%, P < 0.001). The efficiency of 82Rb production improved for both systems over the respective periods of clinical use. CONCLUSIONS: 82Rb generator yield was significantly under-estimated using the CardioGen-82® vs RUBY-FILL® daily QA procedure. When generator yield was expressed as the integrated total activity for both systems, the estimated 82Rb production efficiency of the CardioGen-82® system was ~ 7% lower than RUBY-FILL® over the full period of clinical use.

Comparison of 99mTc Injected Activity with Prescribed Activity in Four Types of Nuclear Medicine Exams.

BACKGROUND: 99mTc is a radioactive isotope that is obtained by eluting a 99Mo/99mTc generator. (PINSTECH, Islamabad) and used for radionuclide scanning. OBJECTIVES: The objective of this work is to study the uncertainties in 99mTc activity that exist due to time delay between injection preparation and administration to patients, during the process of gamma camera scanning. METHODS: Lead canisters were used for storing elution vials and dose calibrator for measuring 99mTc activity in mCi. The activity of preparing 99mTc injection and its administration to patients were compared with the prescribed values of activity recommended in the Society of Nuclear Medicine procedure guidelines. RESULTS: This study showed that uncertainty in the activity existed in one thyroid patient, 38 bone patients, 5 renal patients and 45 cardiac patients. CONCLUSION: This uncertainty in activity exists due to time delay between injection preparation and administration to patients, as well as due to residual radionuclide that is not injected into patients and remains in the syringe.

University of California Replacement of Cesium Irradiators with Alternative Technologies.

The University of California possesses a large number of Cs irradiators that are used in a wide variety of medical and research applications. The university president made a system-wide decision to reduce the potential threat of malevolent use of Cs by switching wherever feasible to x-ray irradiators over a 3-y period of time. A Radioactive Source Replacement Working Group of involved faculty was formed to study the topic and to make recommendations as to when alternative technologies could offer equivalency. The Working Group concluded that x-ray irradiators could replace Cs irradiators in most applications, with some likely exceptions. They found that the depth dose curve for the 320 kVp x-ray irradiator was found to be nearly identical to that of Cs down to a depth in tissue of 4 cm. It was concluded that x rays (energies ≤320 keV) are more biologically effective than Cs gamma rays, suggesting that lower doses of x rays will be required to achieve the same biological endpoint as Cs gamma rays. A simple conversion factor for equating x-ray effects to Cs effects was not recommended because relative biological effectiveness depends on multiple factors. They concluded that each experiment should be individually calibrated when converting from Cs irradiators to x-ray irradiators. The lessons learned from implementing the project to date have shown the importance of having senior management buy-in, involving the research community in the decision making process and allowing for exceptions where equivalency of Cs to x ray cannot be established.

Can an Off-Nominal Landing by an MMRTG-Powered Spacecraft Induce a Special Region on Mars When No Ice Is Present?

This work aims at addressing whether a catastrophic failure of an entry, descent, and landing event of a Multimission Radioisotope Thermoelectric Generator-based lander could embed the heat sources into the martian subsurface and create a local environment that (1) would temporarily satisfy the conditions for a martian Special Region and (2) could establish a transport mechanism through which introduced terrestrial organisms could be mobilized to naturally occurring Special Regions elsewhere on Mars. Two models were run, a primary model by researchers at the Lawrence Berkeley National Laboratory and a secondary model by researchers at the Jet Propulsion Laboratory, both of which were based on selected starting conditions for various surface composition cases that establish the worst-case scenario, including geological data collected by the Mars Science Laboratory at Gale Crater. The summary outputs of both modeling efforts showed similar results: that the introduction of the modeled heat source could temporarily create the conditions established for a Special Region, but that there would be no transport mechanism by which an introduced terrestrial microbe, even if it was active during the temporarily induced Special Region conditions, could be transported to a naturally occurring Special Region of Mars.

Deciding between an X-Ray and 137Cs Irradiator - It's not just about Energy Spectra.

Irradiators utilizing radioactive cesium-137 (137Cs) or cobalt-60 (60Co) gamma-ray sources have been used for biological applications for many decades. These applications include irradiation of much of the nation's blood supply and radiation biology research. In 2005, the U.S. Nuclear Regulatory Commission was assigned the task of preventing the misuse of radioactive materials by persons with malicious intentions; gamma-ray sources, in particular, were given high priority. This resulted in increased security requirements, including constant surveillance, controlled access and personnel background checks. As a result of such regulations being introduced, organizations considering the purchase of a gamma-ray irradiator for the first time or as a replacement to an existing one due to radioactive decay, are now looking into alternative technologies, primarily an X-ray irradiator. To make an educated decision on whether a particular type of X-ray irradiator is of sufficient equivalency to a particular type of 137Cs irradiator for specific applications, one must rely on relevant published comparison studies from other researchers, or perform the comparison studies on their own. This work focuses on the comparison of the radiation physics aspects of two 137Cs irradiator models and three X-ray irradiator models, for the purpose of determining whether the X-ray irradiator models could validly replace the 137Cs irradiator models for certain applications. Although evaluating the influence of relative biological effectiveness (RBE) differences among irradiators could be part of this study, that has been left for a related publication focused on the theoretical aspects of this topic. These evaluations were performed utilizing 47-g and 120-g tissue-equivalent rodent dosimetry phantoms. Our results indicate that, depending upon the user's dose uncertainty budget and maximum areal density of specimens to be irradiated, the RS 2000 160 kVp X-ray irradiator, X-RAD160 X-ray irradiator or X-RAD320 X-ray irradiator could successfully replace a 137Cs irradiator. Technically, any X-ray irradiator model providing similar irradiation geometry, and average energy similar to or higher than these three X-ray models, could also successfully replace a 137Cs irradiator. The results also reveal that differences in inherent source geometry, field geometry and irradiation geometry can counter some of the influence due to differences in energy spectrum. Our goal is that this publication be used as a guide for other similar studies, providing investigators with information on important details that can make the difference between strong and weak comparison conclusions.

Rapid and automated production of [68Ga]gallium chloride and [68Ga]Ga-DOTA-TATE on a medical cyclotron.

INTRODUCTION: The demand for Gallium-68 (68Ga) for labelling PET radiopharmaceuticals has increased over the past few years. 68Ga is obtained through the decayed parent radionuclide 68Ge using commercial 68Ge/68Ga generators. The principal limitation of commercial 68Ge/68Ga generators is that only a limited and finite quantity of 68Ga (<1.85 GBq at start of synthesis) may be accessed. The focus of this study was to investigate the use of a low energy medical cyclotron for the production of greater quantities of 68Ga and to develop an automated and rapid procedure for processing the product. METHODS: Enriched ZnCl2 was electrodeposited on a platinum backing using a NH4Cl (pH 2-4) buffer. The Zn target was irradiated with GE PETtrace 880 at 35 µA and 14.5 and 12.0 MeV beam energy. The irradiated Zn target was purified using octanol resin on an automated system. RESULTS: Following the described procedure, 68Ga was obtained in 6.30 ±â€¯0.42 GBq after 8.5 min bombardment and with low radionuclidic impurities (66Ga (<0.005%) and 67Ga (<0.09%)). Purification on a single octanol resin gave 82% recovery with resulting [68Ga]GaCl3 obtained in 3.5 mL of 0.2 M HCl. [68Ga]GaCl3 production from irradiation to final product was <45 min. To highlight the utility of the automated procedure, [68Ga]Ga-DOTA-TATE labelling was incorporated to give 1.56 GBq at EOS of the labelled peptide with RCY of >70%. CONCLUSIONS: A straightforward procedure for producing 68Ga on a low energy medical cyclotron was described. Current efforts are focus on high activity production and radiolabelling using solid target produced 68Ga.

Successful Migration from Radioactive Irradiators to X-ray Irradiators in One of the Largest Medical Centers in the US.

This paper summarizes about 9 years of effort by Mount Sinai to successfully migrate completely from radioactive irradiators to x-ray irradiators without compromising patient care or research studies. All the effort by Mount Sinai to permanently remove the risk of malicious use of radioactive materials as Radiological Dispersal Device or dirty bomb is reviewed. Due to the unique characteristics of the cesium chloride (CsCl) used in irradiators, it is especially susceptible to be used as a dirty bombs. Mount Sinai originally had four of such irradiators. To reduce and eventually remove the risk of malicious use of radioactive materials, Mount Sinai in New York City has taken several steps. One of such measures was to harden the radioactive irradiators to make the radioactive materials harder to be stolen for malicious purposes. By increasing the delay time, the local law enforcement agency (LLEA) will have more time to stop the intruder. Another measure taken was to implement enhanced security in facilities having radioactive materials. We collaborated with the National Nuclear Security Administration and used state-of-the-art security equipment such as Biometric Access Control and 24/7 video monitoring. In addition, a remote monitoring system with alarms was installed and connected to LLEA for constant monitoring and possible intervention, if necessary, in a timely manner. The other measure taken was to limit the number of people who have access to such radioactive materials. We adopted a single person operator method and reduced the number of people having access from 145 people to only a few people. The adoption of such measures has reduced the risk significantly; however, the best way to remove the permanent risk of these radioactive materials that may be used for a dirty bomb is to use alternative technology to replace these high-activity radioactive sources. In 2013, Mount Sinai purchased its first x-ray irradiator to investigate the feasibility of using x-ray irradiators instead of cesium irradiators for research purposes for cells and small mice. The results from comparison studies were promising, which led to the decision of permanent migration of all cesium irradiators to x-ray irradiators. As of January 2018, Mount Sinai successfully disposed all its Cs irradiators. At this time, Mount Sinai, as one of the largest health care institutions in NY with about 50,000 employees, has migrated completely to alternative technology and removed the risk of malicious use of radioactive materials permanently.

Pre-feasibility Study for Establishing Radioisotope and Radiopharmaceutical Production Facilities in Developing Countries.

BACKGROUND: A significant number of developing countries have no facilities to produce medical radioisotopes and radiopharmaceuticals. OBJECTIVE: In this paper we show that access to life-saving radioisotopes and radiopharmaceuticals and the geographical distribution of corresponding infrastructure is highly unbalanced worldwide. METHODS: We discuss the main issues which need to be addressed in order to establish the production of radioisotopes and radiopharmaceuticals, which are especially important for developing countries as newcomers in the field. The data was gathered from several sources, including databases maintained by the International Atomic Energy Agency (IAEA), World Health Organization (WHO), and other international organizations; personal interactions with representatives in the nuclear medicine field from different regions of the world; and relevant literature. RESULTS: Developing radioisotope and radiopharmaceutical production program and installing corresponding infrastructure requires significant investments, both man-power and financial. Support already exists to help developing countries establish their medical radioisotope production installations from several organizations, such as IAEA. CONCLUSION: This work clearly shows that access to life-saving radioisotopes and the geographical distribution of corresponding infrastructure is highly unbalanced. Technology transfer is important as it not only immediately benefits patients, but also provides employment, economic activity and general prosperity in the region to where the technology transfer is implemented.

Single vial cold kits optimized for preparation of gastrin releasing peptide receptor (GRPR)-radioantagonist 68Ga-RM2 using three different 68Ge/68Ga generators.

68Ga-RM2 is a gastrin releasing peptide receptor (GRPR) antagonist PET (positron emission tomography) radiotracer which is being investigated in clinical trials as a potential prostate cancer imaging agent. Simple, one-step kit formulation of 68Ga-RM2 would facilitate multicentre trials and allow easier integration in hospital radiopharmacy. Herein we report development of three sets of single-vial RM2 cold kits validated for formulation with three respective 68Ge/68Ga generators eluted in 0.6 M, 0.1 M and 0.05 M HCl (hydrochloric acid). Cold kits of varied pH (2, 3, 4 and 5) were prepared using 2 M sodium acetate for three different 68Ge/68Ga generators to determine influence of pH on the radiochemical yield of 68Ga-RM2. Buffer content was optimized with respect to volume of 68GaCl3 eluate to be added (1 mL/2 mL/ 5 mL). Sterility, apyrogenicity and long term stability of cold kits; in vitro and serum stability of 68Ga-RM2 were investigated. In vitro cellular uptake and inhibition studies were performed to demonstrate the specificity of kit-formulated 68Ga-RM2. The radiochemical yield of 68Ga-RM2 formulated from three different generators was observed to be maximum at pH 3 (99 ± 0.5%). Cold kits stored for 6 months at 0 °C also resulted in high radiochemical yield. 68Ga-RM2 exhibited excellent in vitro stability (1 h) and serum stability (1 h). In vitro cellular uptake of 5 ± 0.8% in PC3 cells with >85% inhibition was observed for the 68Ga-RM2 radiotracer indicating its specificity towards GRPR expression. These simple, robust kits shall allow hospitals with different generators to participate in clinical studies of 68Ga-RM2 for screening of GRPR-expressing prostate tumors.

Evaluation of 72Se/72As generator and production of 72Se for supplying 72As as a potential PET imaging radionuclide.

Positron-emitting 72As is the PET imaging counterpart for beta-emitting 77As. Its parent, no carrier added (n.c.a.) 72Se, was produced for a 72Se/72As generator by irradiating an enriched 7°Ge metal-graphite target via the 70Ge(α, 2 n)72Se reaction. Target dissolution used a fast, environmentally friendly method with 93% radioactivity recovery. Chromatographic parameters of the 72Se/72As generator were evaluated, the eluted n.c.a. 72As was characterized with a phantom imaging study, and the previously reported trithiol and aryl-dithiol ligand systems were radiolabeled with the separated n.c.a. 72As in high yield.

Concentration protocol of rhenium-188 perrhenate eluted from tungsten-188/rhenium-188 generator for the preparation of high-yield rhenium-188-labelled radiopharmaceuticals.

Rhenium-188 (Re) is a ß, γ emitter and considered a theranostic radionuclide. It is used for bone pain palliation, treatment of unresectable hepatocellular carcinoma, skin keloids, etc. Re perrhenate is eluted from a W/Re generator in large volumes (8-14 ml) of 0.9% normal saline. Concentrating Re to 1-2 ml volume is important for high radiolabelling yield of various Re radiopharmaceuticals, especially when the generator is old. For this, ion exchange column was prepared in-house, and Re was concentrated using silver ion exchange column and QMA cartridge. Re perrhenate eluted in saline was concentrated to 1 ml with more than 99% yield.

NorthStar Perspectives for Actinium-225 Production at Commercial Scale.

OBJECTIVE: Actinium-225, and its daughter Bismuth-213, have great promise in Alpha Immuno Therapy (AIT) for treatment of various disease modalities. Unfortunately, current production levels of actinium-225 do not support broad use of either actinium-225 or bismuth-213 in development or use for disease treatment. Further, the current cost per millicurie is much too high to be sustainable long term. Resolution of both supply and cost issues allows clinical research to proceed through clinical trials and potentially produce one or more effective therapies for cancer or infectious diseases that could benefit the public. METHODS: NorthStar Medical Technologies, LLC, has investigated several routes that could lead to commercial scale production of actinium-225. RESULTS: This article will discuss those efforts and results to date. CONCLUSION: The outlook for future supplies of actinium-225 from multiple sources to support clinical needs is encouraging.