Federico Zagni

Attenuation Correction for Small Animal PET Images: A Comparison of Two Methods

In order to extract quantitative parameters from PET images, several physical effects such as photon attenuation, scatter, and partial volume must be taken into account. The main objectives of this work were the evaluation of photon attenuation in small animals and the implementation of two attenuation correction methods based on X-rays CT and segmentation of emission images. The accuracy of the first method with respect to the beam hardening effect was investigated by using Monte Carlo simulations. Mouse- and rat-sized phantoms were acquired in order to evaluate attenuation correction in terms of counts increment and recovery of uniform activity concentration. Both methods were applied to mice and rat images acquired with several radiotracers such as18F-FDG, 11C-acetate, 68Ga-chloride, and 18F-NaF. The accuracy of the proposed methods was evaluated in heart and tumour tissues using 18F-FDG images and in liver, kidney, and spinal column tissues using 11C-acetate, 68Ga-chloride, and 18F-NaF images, respectively. In vivo results from animal studies show that, except for bone scans, differences between the proposed methods were about 10% in rats and 3% in mice. In conclusion, both methods provide equivalent results; however, the segmentation-based approach has several advantages being less time consuming and simple to implement.

Accurate modeling of a DOI capable small animal PET scanner using GATE

In this work we developed a Monte Carlo (MC) model of the Sedecal Argus pre-clinical PET scanner, using GATE (Geant4 Application for Tomographic Emission). This is a dual-ring scanner which features DOI compensation by means of two layers of detector crystals (LYSO and GSO). Geometry of detectors and sources, pulses readout and selection of coincidence events were modeled with GATE, while a separate code was developed in order to emulate the processing of digitized data (for example, customized time windows and data flow saturation), the final binning of the lines of response and to reproduce the data output format of the scanners acquisition software. Validation of the model was performed by modeling several phantoms used in experimental measurements, in order to compare the results of the simulations. Spatial resolution, sensitivity, scatter fraction, count rates and NECR were tested. Moreover, the NEMA NU-4 phantom was modeled in order to check for the image quality yielded by the model. Noise, contrast of cold and hot regions and recovery coefficient were calculated and compared using images of the NEMA phantom acquired with our scanner. The energy spectrum of coincidence events due to the small amount of (176)Lu in LYSO crystals, which was suitably included in our model, was also compared with experimental measurements. Spatial resolution, sensitivity and scatter fraction showed an agreement within 7%. Comparison of the count rates curves resulted satisfactory, being the values within the uncertainties, in the range of activities practically used in research scans. Analysis of the NEMA phantom images also showed a good agreement between simulated and acquired data, within 9% for all the tested parameters. This work shows that basic MC modeling of this kind of system is possible using GATE as a base platform; extension through suitably written customized code allows for an adequate level of accuracy in the results. Our careful validation against experimental data confirms that the developed simulation setup is a useful tool for a wide range of research applications.

In-house cyclotron production of high-purity Tc-99m and Tc-99m radiopharmaceuticals

In the last years, the technology for producing the important medical radionuclide technetium-99m by cyclotrons has become sufficiently mature to justify its introduction as an alternative source of the starting precursor [99mTc][TcO4]- ubiquitously employed for the production of 99mTc-radiopharmaceuticals in hospitals. These technologies make use almost exclusively of the nuclear reaction 100Mo(p,2n)99mTc that allows direct production of Tc-99m. In this study, it is conjectured that this alternative production route will not replace the current supply chain based on the distribution of 99Mo/99mTc generators, but could become a convenient emergency source of Tc-99m only for in-house hospitals equipped with a conventional, low-energy, medical cyclotron. On this ground, an outline of the essential steps that should be implemented for setting up a hospital radiopharmacy aimed at the occasional production of Tc-99m by a small cyclotron is discussed. These include (1) target production, (2) irradiation conditions, (3) separation/purification procedures, (4) terminal sterilization, (5) quality control, and (6) Mo-100 recovery. To address these issues, a comprehensive technology for cyclotron-production of Tc-99m, developed at the Legnaro National Laboratories of the Italian National Institute of Nuclear Physics (LNL-INFN), will be used as a reference example.

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.

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, β+: 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 products activity was measured using a calibrated activity meter and a High Purity Germanium gamma-ray detector. RESULTS The saturation yield of68Ga amounts to (330 ± 20) MBq/µA, corresponding to a produced activity of68Ga at the EOB of (4.3 ± 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.

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.

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.

MODELING OF HIP PROSTHESES FOR USE IN RADIOTHERAPY PLANNING: VALIDATION OF A PHOTOGRAMMETRY TECHNIQUE

Treatment planning for external beams radiotherapy for patients with metallic hip prosthesis is often difficult, in particular with regards to the contouring process, in case of poor CT imaging. This may result in inaccurate dosimetry for the femoral bone and regions next to it. In this work an alternative method to obtain more accurate prosthetic contourings, based on photogrammetry, was developed. More precisely, the process of 3D modeling of a femoral prosthetic component and the conversion to DICOM was developed, plus the effective employment in a treatment planning system was evaluated. Validation of the obtained model was performed by comparing several cross sections against experimental measurements. Comparison resulted in very low deviations, showing good accuracy of the photogrammetry technique, despite the high asymmetry of the object. The tests conducted in a research TPS indicate that models obtained with the proposed method are able to provide additional information in order to better characterize such prosthetic implants.