Production of Ga-68 with a General Electric PETtrace cyclotron by liquid target.

PURPOSE: In recent years the use of 68Ga (t1/2 = 67.84 min, ß+: 88.88%) for the labelling of different PET radiopharmaceuticals has significantly increased. This work aims to evaluate the feasibility of the production of 68Ga via the 68Zn(p,n)68Ga reaction by proton irradiation of an enriched zinc solution, using a biomedical cyclotron, in order to satisfy its increasing demand. METHODS: Irradiations of 1.7 Msolution of 68Zn(NO3)2 in 0.2 N HNO3 were conducted with a GE PETtrace cyclotron using a slightly modified version of the liquid target used for the production of fluorine-18. The proton beam energy was degraded to 12 MeV, in order to minimize the production of 67Ga through the68Zn(p,2n)67Ga reaction. The product's activity was measured using a calibrated activity meter and a High Purity Germanium gamma-ray detector. RESULTS: The saturation yield of68Ga amounts to (330 ±â€¯20) MBq/µA, corresponding to a produced activity of68Ga at the EOB of (4.3 ±â€¯0.3) GBq in a typical production run at 46 µA for 32 min. The radionuclidic purity of the68Ga in the final product, after the separation, is within the limits of the European Pharmacopoeia (>99.9%) up to 3 h after the EOB. Radiochemical separation up to a yield not lower than 75% was obtained using an automated purification module. The enriched material recovery efficiency resulted higher than 80-90%. CONCLUSIONS: In summary, this approach provides clinically relevant amounts of68Ga by cyclotron irradiation of a liquid target, as a competitive alternative to the current production through the68Ge/68Ga generators.

Radiation Protection Studies for Medical Particle Accelerators using Fluka Monte Carlo Code.

Radiation protection (RP) in the use of medical cyclotrons involves many aspects both in the routine use and for the decommissioning of a site. Guidelines for site planning and installation, as well as for RP assessment, are given in international documents; however, the latter typically offer analytic methods of calculation of shielding and materials activation, in approximate or idealised geometry set-ups. The availability of Monte Carlo (MC) codes with accurate up-to-date libraries for transport and interaction of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of modern computers, makes the systematic use of simulations with realistic geometries possible, yielding equipment and site-specific evaluation of the source terms, shielding requirements and all quantities relevant to RP at the same time. In this work, the well-known FLUKA MC code was used to simulate different aspects of RP in the use of biomedical accelerators, particularly for the production of medical radioisotopes. In the context of the Young Professionals Award, held at the IRPA 14 conference, only a part of the complete work is presented. In particular, the simulation of the GE PETtrace cyclotron (16.5 MeV) installed at S. Orsola-Malpighi University Hospital evaluated the effective dose distribution around the equipment; the effective number of neutrons produced per incident proton and their spectral distribution; the activation of the structure of the cyclotron and the vault walls; the activation of the ambient air, in particular the production of 41Ar. The simulations were validated, in terms of physical and transport parameters to be used at the energy range of interest, through an extensive measurement campaign of the neutron environmental dose equivalent using a rem-counter and TLD dosemeters. The validated model was then used in the design and the licensing request of a new Positron Emission Tomography facility.

Assessment of the neutron dose field around a biomedical cyclotron: FLUKA simulation and experimental measurements.

In the planning of a new cyclotron facility, an accurate knowledge of the radiation field around the accelerator is fundamental for the design of shielding, the protection of workers, the general public and the environment. Monte Carlo simulations can be very useful in this process, and their use is constantly increasing. However, few data have been published so far as regards the proper validation of Monte Carlo simulation against experimental measurements, particularly in the energy range of biomedical cyclotrons. In this work a detailed model of an existing installation of a GE PETtrace 16.5MeV cyclotron was developed using FLUKA. An extensive measurement campaign of the neutron ambient dose equivalent H∗(10) in marked positions around the cyclotron was conducted using a neutron rem-counter probe and CR39 neutron detectors. Data from a previous measurement campaign performed by our group using TLDs were also re-evaluated. The FLUKA model was then validated by comparing the results of high-statistics simulations with experimental data. In 10 out of 12 measurement locations, FLUKA simulations were in agreement within uncertainties with all the three different sets of experimental data; in the remaining 2 positions, the agreement was with 2/3 of the measurements. Our work allows to quantitatively validate our FLUKA simulation setup and confirms that Monte Carlo technique can produce accurate results in the energy range of biomedical cyclotrons.

Experimental measurement and Monte Carlo assessment of Argon-41 production in a PET cyclotron facility.

In a medical cyclotron facility, (41)Ar (t1/2 = 109.34 m) is produced by the activation of air due to the neutron flux during irradiation, according to the (40)Ar(n,γ)(41)Ar reaction; this is particularly relevant in widely diffused high beam current cyclotrons for the production of PET radionuclides. While theoretical estimations of the (41)Ar production have been published, no data are available on direct experimental measurements for a biomedical cyclotron. In this work, we describe a sampling methodology and report the results of an extensive measurement campaign. Furthermore, the experimental results are compared with Monte Carlo simulations performed with the FLUKA code. To measure (41)Ar activity, air samples were taken inside the cyclotron bunker in sealed Marinelli beakers, during the routine production of (18)F with a 16.5 MeV GE-PETtrace cyclotron; this sampling thus reproduces a situation of absence of air changes. Samples analysis was performed in a gamma-ray spectrometry system equipped with HPGe detector. Monte Carlo assessment of the (41)Ar saturation yield was performed directly using the standard FLUKA score RESNUCLE, and off-line by the convolution of neutron fluence with cross section data. The average (41)Ar saturation yield per one liter of air of (41)Ar, measured in gamma-ray spectrometry, resulted to be 3.0 ± 0.6 Bq/µA*dm(3) while simulations gave a result of 6.9 ± 0.3 Bq/µA*dm(3) in the direct assessment and 6.92 ± 0.22 Bq/µA*dm(3) by the convolution neutron fluence-to-cross section.

Synthesis and quality control of 68Ga citrate for routine clinical PET.

INTRODUCTION AND AIM: Scintigraphic imaging of infection and inflammation with 67Ga-citrate is an established and powerful diagnostic tool in the management of patients with infectious or inflammatory diseases. 68Ga is a short-lived positron-emitting radionuclide (half-life 67.6 min, positron energy 2.92 MeV), which allows better imaging qualities than 67Ga using the high spatial resolution and the quantitative features of PET. The aim of this study was to develop a method of synthesis for 68Ga citrate with high and reproducible radiochemical yield using a commercial 68Ga-labelling module. The resultant 68Ga citrate would be suitable for use in the detection of infectious or inflammatory diseases in routine clinical practice. METHODS: A simplified method of producing 68Ga citrate is described. Radiochemical purity, pyrogen testing were performed as per the standard protocols. RESULTS: After performing 10 syntheses of 68Ga citrate, the radiochemical yield was 64.1+/-6.0% (mean+/-standard deviation) with an average activity of 971.2+/-103.4 MBq available for labelling. Radiochemical purity determined by instant thin-layer chromatography-silica gel was higher than 98%. All the synthesized products were found to be sterile and pyrogen-free. In this study, the quality control step provided good and reproducible results. This is worth noting, especially in view of the stringent new rules adopted in most European countries for the in-house good manufacturing practice (GMP) synthesis of radiopharmaceuticals. CONCLUSION: The high radiochemical yield and purity showed that this method is a reliable tool for the production of 68Ga citrate to be used in the detection of inflammatory and infectious diseases using high resolution and qualitative PET.

Radiation emission dose from patients administered 90Y-labelled radiopharmaceuticals: comparison of experimental measurements versus Monte Carlo simulation.

AIM: To estimate the radiation dose delivered from patients injected with yttrium-90 (Y)-labelled tiuxetan (Zevalin) to parents and the general population, comparing different techniques. METHODS: The radiation dose delivered from a group of eight patients injected with Y-Zevalin to treat recurrent lymphoma was measured. The data obtained with the Monte Carlo simulation test were compared with the experimental measurements obtained with an ionization chamber detector and with a crystal NaI(Tl) detector. RESULTS: A good correlation was found between the Monte Carlo simulation test and the ionization chamber detector results: the air kerma dose rate was 4.2+/-0.1 and 4.4+/-0.8 microGy/h, respectively (r=0.9, P<0.01). Moreover, more than 99.7% of the air kerma dose rate measured with the ionization chamber detector was because of the contribution of electrons, whereas the contribution of photons was less than 0.3%. In contrast, the air kerma dose rate measured with the crystal NaI(Tl) detector was significantly lower (0.76+0.12 microGy/h) in comparison with the Monte Carlo simulation test. This underestimation was related to the limited crystal NaI(Tl) detector response to low energy rates at variance with the ionization chamber detector. The effective radiation dose released by patients treated with Y-labelled tiuxetan to parents and the general population was approximately 0.1 mSv per treatment cycle. CONCLUSION: Using the Monte Carlo model as a benchmark to compare the experimental measurements obtained by the two different detectors, we found that the ionizing chamber detector was more accurate than the crystal Na(Tl) detector for measuring the exposure radiation dose delivered from patients administered with Y-labelled radiopharmaceuticals. Moreover, the effective radiation dose released by these patients to their parents and the general population is significantly lower than the value recommended by international reports and regulations.

Assessment of radionuclidic impurities in 2-[18F]fluoro-2-deoxy-d-glucose ([18F]FDG) routine production.

In this paper, radionuclidic impurities generated during the bombardment of [18 O]water in the routine production of 2-[18F]fluoro-2-deoxy-d-glucose ([18F]FDG) were studied. In order to assess such impurities and the efficacy of purification methods through the different steps of the synthesis, samples of the target filters, purification columns, [18 O]water recovered after the synthesis, and the final solution was collected and their activities measured and analyzed by means of a gamma-ray spectrometry system. The data demonstrated that purification methods adopted for the synthesis provide the [18F]FDG radionuclidically pure, as requested by the EU Pharmacopeia.