Angelo Infantino

Prediction of 89Zr production using the Monte Carlo code FLUKA

The widely used Monte Carlo simulation code FLUKA has been utilized to prototype a solid target for the production of (89)Zr by irradiation of a metallic (89)Y target foil in a 16.5MeV proton biomedical cyclotron, through the reaction (89)Y(p, n)(89)Zr. Simulations were performed with and without an Al energy degrader. In the setup of the geometry of the target, state of the art support tools, like SimpleGeo, were used for accurate, detailed modeling. The results permitted a quick assessment of all possible radionuclidic contaminants and confirmed that the use of an energy degrader avoids production of the most important impurity, (88)Zr. The estimated value for the activity produced in one hour of irradiation at 20μA is 384 ± 42MBq; this is encouraging, indicating possible production of clinically significant amounts of activity with the relatively simple target setup adopted. Initial experimental tests gave results in excellent agreement with simulations, confirming the usefulness and accuracy of FLUKA as a tool for the design and optimization of targets for the production of PET radionuclides.

Some experimental studies on 89Zr production

Abstract The radionuclide 89Zr (T1/2=78.4ߙh) is particularly attractive for in vivo assessment of biochemical proesses characterized by biological half times in the order of several hours. In the present study we assessed the feasibility of 89Zr production via the 89Y(p,n)89Zr reaction. Irradiation tests were performed using a 16.5ߙMeV GE-PETtrace cyclotron. High purity (>99.9%) yttrium metallic foils (0.15ߙmm) were irradiated in a solid target station developed in our Institution. The literature cross section data of all the nuclear reactions possible in the energy range of interest were carefully studied to optimize the irradiation parameters. The irradiated target was dissolved in 1ߙN HCl and the activity of the sample was measured in a CRC-15 PET dose calibrator, setting different values for the calibration factor. The sample activity was then measured usin g a calibrated HPGe gamma ray detector. A comparison of the activity measurements allowed the evaluation of the calibration factor for the Capintec CRC-15 PET; for a 3ߙml syringe geometry, it was found 739±60. Saturation yields of 89Zr were found to be 1150±110ߙMBq/μA for a single 0.15ߙmm thick disc irradiation (Ep=12.6→11.2ߙMeV) and 2400±220ߙMBq/μA for irradiation of two discs of total thickness 0.3ߙmm (Ep=12.6→9.5MeV). Total produced activities in the range of 200–400ߙMBq (according to target thickness) were repeatedly and safely obtained by 60ߙmin bombardments at 20ߙμA. Analysis of gamma ray spectra of all the samples showed a very high radionuclidic purity (>99.9995%).

Monte Carlo Evaluation of Single Event Effects in a Deep-Submicron Bulk Technology: Comparison Between Atmospheric and Accelerator Environment

In this work, the expected SEE rate in a generic model of a state-of-the-art SRAM memory was studied as a function of the critical charge in different radiation environments. The FLUKA Monte Carlo code was used to evaluate the SEE mono-energetic cross section for different particles as well as the SEE rate in different mixed field environments through an energy deposition analysis. Mono-energetic cross sections for protons and both positively and negatively charged muons were evaluated in the 0.1–1000 MeV energy range. The SEE rate was calculated for different mixed radiation field environments such as the terrestrial environment, commercial flight altitude, LHC critical areas for electronics and CERN’s CHARM test facility. Results show that direct ionization from charged particles becomes predominant between 0.2-1.1 fC compared to indirect neutron energy deposition. Finally the CHARM facility was demonstrated to be able to reproduce with a good approximation both the atmospheric and accelerator environments, particularly with regard to avionics, ground level applications and LHC.

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.

Characterization of 41Ar production in air at a PET cyclotron facility

In the production of Positron Emission Tomography (PET) nuclides at a medical cyclotron facility 41Ar (T1/2 = 109.34 m) is produced by the activation of air due to the neutron flux, according to the 40Ar(n, γ)41Ar reaction. In this work, we describe a relatively inexpensive and readily reproducible methodology of air sampling that can be used for quantification of 41Ar during the routine production of PET nuclides. We report the results of an extensive measurement campaign in the cyclotron bunker and in the ducts of the ventilation system, before and after final filtering of the extracted air. Air Samples were analyzed using a gamma-ray spectrometry system equipped with HPGe detector, with proper correction of the efficiency calibration to account for the samples density. The results of measurement were then used to evaluate the Total Effective Dose (TED) to the population living in the surrounding areas, due to routine emissions in the operation of the cyclotron. The average 41Ar saturation yield per one liter of air emitted in the environment resulted to be (0.044 ± 0.007) Bq/(μA ⋅ dm3). The maximum value of TED for the critical group of the population, even considering an overestimated workload, was less than 0.19 μSv/year, well below the level of radiological relevance.

Efficiency calibration of a portable CZT detector for nondestructive activation assessment of a cyclotron bunker

ABSTRACT During the operational life of a PET Cyclotron, the concrete walls of the cyclotron vault are activated by the secondary neutron flux interacting with rare earths and metals present in the concrete or in reinforcement bars. For this reason when considering dismantling of such accelerators, the amount of radioactive waste has to be evaluated in advance to identify any critical issues or possible countermeasures to be taken to define an optimum decommissioning strategy. The aim of this work is to define a non-destructive in situ measurement methodology for a preliminary activation assessment of a cyclotron bunker with no need for core drilling. A very compact, USB-powered, CdZnTe (CZT) detector for gamma-ray spectrometry was used for the activation assessment of the site of installation of a GE PETtrace (16.5 MeV) cyclotron, routinely used in the production of positron-emitting radionuclides. Because of the complexity of measurement geometry, the efficiency calibration of the detector was performed via Monte Carlo (MC) simulations. The detector was accurately modelled using FLUKA, including a 5 cm lead shielding set-up. The MC model of the detector was validated for a wide range of energies and different source geometries, showing discrepancies below 5% for all tested sources. The efficiency curve for wall activation measurements was calculated, allowing a quantitative evaluation of activity concentration.

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.

Undesired radionuclides in 18F2 production by deuterons

In this paper, radionuclidic impurities generated during the bombardment of an 20Ne gaseous matrix in the routine production of [18F]-fluorodopa were studied. 18F2 is produced by the GE PETtrace cyclotron capable of accelerating negative deuteron ions to an energy of 8.4 MeV. The materials of the target and of the Havar foil, the stable fluorine added as carrier and the natural isotope composition of neon may interact with deuterons and produce undesired radionuclides. To measure them, a sampling system has been realized. Samples of the irradiated gas have been measured by gamma-ray spectrometry. The results show how the radionuclidic purity obtained is very high after a short time from the end of bombardment (EOB).

Modeling of a Cyclotron Target for the Production of 11C with Geant4

BACKGROUND In medical cyclotron facilities, 11C is produced according to the 14N(p,α)11C reaction and widely employed in studies of prostate and brain cancers by Positron Emission Tomography. It is known from literature that the 11C-target assembly shows a reduction in efficiency during time, meaning a decrease of activity produced at the end of bombardment. This effect might depend on aspects which are still not completely known. OBJECTIVE Possible causes of the loss of performance of the 11C-target assembly were addressed by Monte Carlo simulations. METHODS Geant4 was used to model the 11C-target assembly of a GE PETtrace cyclotron. The physical and transport parameters to be used in the energy range of medical applications were extracted from literature data and 11C routine productions. The Monte Carlo assessment of 11C saturation yield was performed varying several parameters such as the proton energy and the angle of the target assembly with respect to the proton beam. RESULTS The estimated 11C saturation yield is in agreement with IAEA data at the energy of interest, while it is about 35% greater than the experimental value. A more comprehensive modeling of the target system, including thermodynamic effect, is required. The energy absorbed in the inner layer of the target chamber was up to 46.5 J/mm2 under typical irradiation conditions. CONCLUSION This study shows that Geant4 is potentially a useful tool to design and optimize targetry for PET radionuclide productions. Tests to choose the Geant4 physics libraries should be performed before using this tool with different energies and materials.

Validating production of PET radionuclides in solid and liquid targets: Comparing Geant4 predictions with FLUKA and measurements

The Monte Carlo toolkit Geant4 is used to simulate the production of a number of positron emitting radionuclides: 13N, 18F, 44Sc, 52Mn, 55Co 61Cu, 68Ga, 86Y, 89Zr and 94Tc, which have been produced using a 13MeV medical cyclotron. The results are compared to previous simulations with the Monte Carlo code FLUKA and experimental measurements. The comparison shows variable degrees of agreement for different isotopes. The mean absolute deviation of Monte Carlo results from experiments was 1.4±1.6 for FLUKA and 0.7±0.5 for Geant4 using TENDL cross sections with QGSP-BIC-AllHP physics. Both agree well within the large error, which is due to the uncertainties present in both experimentally determined and theoretical reaction cross sections. Overall, Geant4 has been confirmed as a tool to simulate radionuclide production at low proton energy.