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.

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.

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.

Investigation of Workplace-like Calibration Fields via a Deuterium-Tritium (D-T) Neutron Generator.

Radiation survey meters and personal dosimeters are typically calibrated in reference neutron fields based on conventional radionuclide sources, such as americium-beryllium (Am-Be) or californium-252 (Cf), either unmodified or heavy-water moderated. However, these calibration neutron fields differ significantly from the workplace fields in which most of these survey meters and dosimeters are being used. Although some detectors are designed to yield an approximately dose-equivalent response over a particular neutron energy range, the response of other detectors is highly dependent upon neutron energy. This, in turn, can result in significant over- or underestimation of the intensity of neutron radiation and/or personal dose equivalent determined in the work environment. The use of simulated workplace neutron calibration fields that more closely match those present at the workplace could improve the accuracy of worker, and workplace, neutron dose assessment. This work provides an overview of the neutron fields found around nuclear power reactors and interim spent fuel storage installations based on available data. The feasibility of producing workplace-like calibration fields in an existing calibration facility has been investigated via Monte Carlo simulations. Several moderating assembly configurations, paired with a neutron generator using the deuterium tritium (D-T) fusion reaction, were explored.

A simple method for preparation of pure 68 Ga-acetate precursor for formulation of radiopharmaceuticals: Physicochemical characteristics of the 68 Ga eluate of the SnO2 based-68 Ge/68 Ga column generator.

Gallium-68 radioisotope is an excellent source in clinical positron emission tomography application due to its ease of availability from germanium-68 (68 Ge)/gallium-68 (68 Ga) generator having a shelf life of 1 year. In this paper, a modified method for purification of the primary eluate of 68 Ge-68 Ga generator by using a small cation exchange resin (Dowex-50) column has been described. The breakthrough of 68 Ge before and after purification of 68 Ga eluate was 0.014% and 0.00027%, respectively. The average recovery yield of 68 Ga after purification was 84% ± 8.6% (SD, n = 335). The results of the physiochemical studies confirmed that the 68 Ga-acetate obtained is suitable for labeling of radiopharmaceuticals.

Fission-Produced 99Mo Without a Nuclear Reactor.

99Mo, the parent of the widely used medical isotope 99mTc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of 99Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of 99Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The 99Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq 99Mo production run are presented.

Comprehensive Quality Control of the ITG 68Ge/68Ga Generator and Synthesis of 68Ga-DOTATOC and 68Ga-PSMA-HBED-CC for Clinical Imaging.

UNLABELLED: A good-manufacturing-practices (GMP) (68)Ge/(68)Ga generator that uses modified dodecyl-3,4,5-trihydroxybenzoate hydrophobically bound to a octadecyl silica resin (C-18) as an adsorbent has been developed that allows for dilute HCl (0.05N) to efficiently elute metal-impurity-free (68)Ga(3+) ready for peptide labeling. We characterized the performance of this generator system over a year in conjunction with the production of (68)Ga-labeled DOTATOC and Glu-NH-CO-NH-Lys(Ahx)-HBED-CC (PSMA-HBED-CC) intended for clinical studies and established protocols for batch release. METHODS: A 2,040-MBq self-shielded (68)Ge/(68)Ga generator provided metal-free (68)GaCl3 ready for peptide labeling in the fluidic labeling module after elution with 4 mL of 0.05N HCl. The compact system was readily housed in a laminar flow cabinet allowing an ISO class-5 environment. (68)Ga labeling of peptides using GMP kits was performed in 15-20 min, and the total production time was 45-50 min. Batch release quality control specifications were established to meet investigational new drug submission and institutional review board approval standards. RESULTS: Over a period of 12 mo, (68)Ga elution yields from the generator averaged 80% (range, 72.0%-95.1%), and (68)Ge breakthrough was less than 0.006%, initially decreasing with time to 0.001% (expressed as percentage of (68)Ge activity present in the generator at the time of elution), a unique characteristic of this generator. The radiochemical purity of both (68)Ga-DOTATOC and (68)Ga-PSMA-HBED-CC determined by high-performance liquid chromatography analysis was greater than 98%, with a minimum specific activity of 12.6 and 42 GBq/µmol, respectively. The radionuclidic ((68)Ge) impurity was 0.00001% or less (under the detection limit). Final sterile, pyrogen-free formulation was provided in physiologic saline with 5%-7% ethanol. CONCLUSION: The GMP-certified (68)Ge/(68)Ga generator system was studied for a year. The generator system is contained within the fluidic labeling module, and it is compact, self-shielded, and easy to operate using simple manual techniques. The system provides radiolabeled peptides with high (>98%) radiochemical purity and greater than 80% radiochemical yield. The (68)Ge levels in the final drug products were under the detection limits at all times. (68)Ga-DOTATOC and (68)Ga-PSMA-HBED-CC investigational radiopharmaceuticals are currently being studied clinically under investigational new drug (IND) applications submitted to the U.S. Food and Drug Administration.

Fractionated elution using the TEKCIS technetium-99m generator.

The TEKCIS technetium-99m (Tc) generator was designed to allow dry column shipment and automatized conception. A high Tc radioactive concentration is required in a subset of radiopharmacy procedures. Fractionated elution can be a useful tool to meet this requirement, especially when current elution is close to the generator expiration date. The aim of our study was to assess TEKCIS generator elution kinetics and to determine the optimal fractionated elution time to fit with procedures requiring the highest Tc radioactive concentration in clinical use. After duplicate elution at several predetermined elution times, the volume and activity of each eluate were measured. Two optimal time points were selected to perform fractionated elution and repeatability (n=34 and 33) assessed on TEKCIS generators calibrated at 6 or 8 GBq. The complete eluate volume (5 ml) was collected after 60 s of elution. A logarithmic equation was established between eluate volume (v, ml) from elapsed elution time (t, s): v=1.8335ln(t)-2.5965. Using the reciprocal equation, elution times required to obtain some commonly eluted volumes were calculated. Fractionated elutions during 15 and 20 s were selected and an average elution volume from 2.74 to 3.27 ml was collected, with an average elution yield of approximately 90 and 100%, respectively. Our work provides a simple and reliable methodology for the use of fractionated elution with the new TEKCIS generator.

Detailed evaluation of different (68)Ge/(68)Ga generators: an attempt toward achieving efficient (68)Ga radiopharmacy.

The present study is aimed at carrying out a comparative performance evaluation of different types of (68)Ge/(68)Ga generators to identify the best choice for use in (68)Ga-radiopharmacy. Over the 1 year period of evaluation, the elution yields from the CeO2-based and SiO2-based (68)Ge/(68) Ga generators remained almost consistent, in contrast to the sharp decrease observed in the elution yields from TiO2 and SnO2-based generators. The level of (68)Ge impurity in (68)Ga eluates from the CeO2 and SiO2-based (68)Ge/(68)Ga generator was always <10(-3)%, while this level increased from 10(-3)% to 10(-1)% in case of TiO2 and SnO2-based generators. The level of chemical impurities in (68)Ga eluates from CeO2 and SiO2-based (68)Ge/(68)Ga generators was negligibly low (<0.1 ppm) in contrast to the significantly higher level (1-20 ppm) of such impurities in eluates from other two generators. As demonstrated by radiolabeling studies carried out using DOTA-coupled dimeric cyclic RGD peptide derivative (DOTA-RGD2), CeO2-PAN and SiO2-based generators are directly amenable for radiopharmaceutical preparation, whereas the other generators can be only used after post-elution purification of (68)Ga eluates. Clinically relevant dose of (68)Ga-DOTA-RGD2 was prepared in a hospital radiopharmacy for non-invasive visualization of tumors in breast cancer patients using positron emission tomography.

Exploration of Adiabatic Resonance Crossing Through Neutron Activator Design for Thermal and Epithermal Neutron Formation in (99)Mo Production and BNCT Applications.

A feasibility study was performed to design thermal and epithermal neutron sources for radioisotope production and boron neutron capture therapy (BNCT) by moderating fast neutrons. The neutrons were emitted from the reaction between (9)Be, (181)Ta, and (184)W targets and 30 MeV protons accelerated by a small cyclotron at 300 µA. In this study, the adiabatic resonance crossing (ARC) method was investigated by means of (207)Pb and (208)Pb moderators, graphite reflector, and boron absorber around the moderator region. Thermal/epithermal flux, energy, and cross section of accumulated neutrons in the activator were examined through diverse thicknesses of the specified regions. Simulation results revealed that the (181)Ta target had the highest neutron yield, and also tungsten was found to have the highest values in both surface and volumetric flux ratio. Transmutation in the (98)Mo sample through radiative capture was investigated for the natural lead moderator. When the sample radial distance from the target was increased inside the graphite region, the production yield had the greatest value of activity. The potential of the ARC method is a replacement or complements the current reactor-based supply sources of BNCT purposes.

Fully-automated synthesis of 16ß-(18)F-fluoro-5α-dihydrotestosterone (FDHT) on the ELIXYS radiosynthesizer.

Noninvasive in vivo imaging of androgen receptor (AR) levels with positron emission tomography (PET) is becoming the primary tool in prostate cancer detection and staging. Of the potential (18)F-labeled PET tracers, (18)F-FDHT has clinically shown to be of highest diagnostic value. We demonstrate the first automated synthesis of (18)F-FDHT by adapting the conventional manual synthesis onto the fully-automated ELIXYS radiosynthesizer. Clinically-relevant amounts of (18)F-FDHT were synthesized on ELIXYS in 90 min with decay-corrected radiochemical yield of 29±5% (n=7). The specific activity was 4.6 Ci/µmol (170 GBq/µmol) at end of formulation with a starting activity of 1.0 Ci (37 GBq). The formulated (18)F-FDHT yielded sufficient activity for multiple patient doses and passed all quality control tests required for routine clinical use.

Diversification of 99Mo/99mTc separation: non­fission reactor production of 99Mo as a strategy for enhancing 99mTc availability.

This paper discusses the benefits of obtaining (99m)Tc from non-fission reactor-produced low-specific-activity (99)Mo. This scenario is based on establishing a diversified chain of facilities for the distribution of (99m)Tc separated from reactor-produced (99)Mo by (n,γ) activation of natural or enriched Mo. Such facilities have expected lower investments than required for the proposed chain of cyclotrons for the production of (99m)Tc. Facilities can receive and process reactor-irradiated Mo targets then used for extraction of (99m)Tc over a period of 2 wk, with 3 extractions on the same day. Estimates suggest that a center receiving 1.85 TBq (50 Ci) of (99)Mo once every 4 d can provide 1.48-3.33 TBq (40-90 Ci) of (99m)Tc daily. This model can use research reactors operating in the United States to supply current (99)Mo needs by applying natural (nat)Mo targets. (99)Mo production capacity can be enhanced by using (98)Mo-enriched targets. The proposed model reduces the loss of (99)Mo by decay and avoids proliferation as well as waste management issues associated with fission-produced (99)Mo.

Cryogenic molecular separation system for radioactive (11)C ion acceleration.

A (11)C molecular production/separation system (CMPS) has been developed as part of an isotope separation on line system for simultaneous positron emission tomography imaging and heavy-ion cancer therapy using radioactive (11)C ion beams. In the ISOL system, (11)CH4 molecules will be produced by proton irradiation and separated from residual air impurities and impurities produced during the irradiation. The CMPS includes two cryogenic traps to separate specific molecules selectively from impurities by using vapor pressure differences among the molecular species. To investigate the fundamental performance of the CMPS, we performed separation experiments with non-radioactive (12)CH4 gases, which can simulate the chemical characteristics of (11)CH4 gases. We investigated the separation of CH4 molecules from impurities, which will be present as residual gases and are expected to be difficult to separate because the vapor pressure of air molecules is close to that of CH4. We determined the collection/separation efficiencies of the CMPS for various amounts of air impurities and found desirable operating conditions for the CMPS to be used as a molecular separation device in our ISOL system.

The radioisotope complex project "RIC-80" at the Petersburg Nuclear Physics Institute.

The high current cyclotron C-80 capable of producing 40-80 MeV proton beams with a current of up to 200 µA has been constructed at Petersburg Nuclear Physics Institute. One of the main goals of the C-80 is the production of a wide spectrum of medical radionuclides for diagnostics and therapy. The project development of the radioisotope complex RIC-80 (radioisotopes at the cyclotron C-80) at the beam of C-80 has been completed. The RIC-80 complex is briefly discussed in this paper. The combination of the mass-separator with the target-ion source device, available at one of the new target stations for on-line or semi on-line production of a high purity separated radioisotopes, is explored in greater detail. The results of target and ion source tests for a mass-separator method for the production of high purity radioisotopes (82)Sr and (223,224)Ra are also presented.

Performance of a 6²Zn/6²Cu microgenerator in kit-based synthesis and delivery of [6²Cu]Cu-ETS for PET perfusion imaging.

The performance of a commercially produced (62)Zn/(62)Cu microgenerator system, and an associated kit-based radiopharmaceutical synthesis method, was evaluated for clinical site production of [(62)Cu]Cu-ETS (ethylglyoxal bis(thiosemicarbazonato)copper(II)), an investigational agent for PET perfusion imaging. Using 37 generators, containing 1.84±0.23 GBq (62)Zn at 9:00 AM on the day of clinical use, a total of 45 patient doses of [(62)Cu]Cu-ETS (672±172 MBq) were delivered without difficulty. (62)Cu elution yields were high (approximately 90%), accompanied by extremely low (62)Zn breakthrough (<0.001%). Radiopharmaceutical preparation, from the start-of-elution to time-of-injection, consumed less than five minutes. The (62)Zn/(62)Cu microgenerator was a dependable source of short-lived positron-emitting (62)Cu, and the kit-based synthesis proved to be rapid, robust, and highly reliable for "on-demand" delivery of [(62)Cu]Cu-ETS for PET perfusion imaging.

Analysis of induced radionuclides in replacement parts and liquid wastes in a medical cyclotron solely used for production of 18F for [18F]FDG.

Radioactivities produced in replacement parts and liquid wastes in a medical cyclotron used to produce (18)F for [(18)F]FDG with 10MeV protons were analyzed. Nineteen radionuclides were found in the replacement parts and liquid wastes. Among them, long-lived (56)Co in the Havar foils is critical in terms of radioactive waste management. The estimated dose level of exposure for the operating staff during the replacement of parts was around 310µSv/y, which is smaller than the recommended dose limit for workers.

Production of 13N by 12C(d,n)13N reaction in a medium energy plasma focus.

This paper explores the production of (13)N by bombardment of a carbon target by high energy deuterons in a medium energy plasma focus. A set of experiments in the energy range of 2.7-3.1kJ and initial pressure of 200-700Pa, with three or five shots in each experiment, was performed. A HPGe detector was used for gamma spectroscopy, and 511keV photons emitted by positron annihilation were utilized to measure the (13)N radioactivity. The highest activity of (13)N in these experiments was 14Bq which was acquired after five shots at a pressure of 450Pa and a 3.1kJ stored energy. Calculations based on thick target yield showed that at least 1.9×10(9) deuterons with energies higher than 330keV were ejected from the pinch region.

82Sr purification procedure using Chelex-100 resin.

(82)Rb is a positron-emitting radionuclide widely used in nuclear cardiology. One great advantage is its availability through a generator loaded with (82)Sr. (82)Sr can be produced in a high energy cyclotron by irradiating rubidium chloride target with proton beam. In this paper, we present an extensive study (elution profiles, effect of the elution flow rate) on the use of Chelex-100 resin and ammonia buffer. No significant effect of flow rate was evidenced between 1 and 10mL/min leading us to propose a purification process which can be easily automated.

Development of a 99Mo/99mTc generator using alumina microspheres for industrial radiotracer applications.

A chromatographic (99)Mo/(99m)Tc generator for industrial applications has been developed using alumina microspheres synthesized through sol-gel process to obtain (99m)Tc in both aqueous and non-aqueous media. The sorbent was mesoporous, mechanically strong and possessed high surface area. (99m)Tc could be eluted from generator system using either acetone or 0.9% NaCl solution with appreciably high yields and high radiochemical as well as radionuclidic purity. The facile, versatile generator provides an efficient way to access (99m)Tc at industrial sites for radiotracer applications.