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In this work a set of calibration constants to measure the 18F airborne activity in straight ducts with square section are computed by means of a Geant4 simulation and provided for different duct lengths. The calibration constants are compared with analytical calculations and experimental measurements. The simulated calibration constants provide a useful cross-check for the actual calibration of airborne monitoring systems in ducts, which is practically complex and can be affected by large uncertainties.

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.

OBJECTIVE: When handling Y-labelled and Lu-labelled radiopharmaceuticals, skin exposure is mainly due to β-particles. This study aimed to investigate the equivalent dose saving of the staff when changing from an essentially manual radiolabelling procedure to an automatic dose dispenser (ADD). MATERIALS AND METHODS: The chemist and physician were asked to wear thermoluminescence dosimeters on their fingertips to evaluate the quantity of Hp(0.07) on the skin. Data collected were divided into two groups: before introducing ADD (no ADD) and after introducing ADD. RESULTS: For the chemist, the mean values (95th percentile) of Hp(0.07) for no ADD and ADD are 0.030 (0.099) and 0.019 (0.076) mSv/GBq, respectively, for Y, and 0.022 (0.037) and 0.007 (0.023) mSv/GBq, respectively, for Lu. The reduction for ADD was significant (t-test with P

This work was aimed at characterizing and validating a very compact, USB-powered, CdZnTe detector for gamma-ray spectrometry applications in radiopharmacy. A GR1 model (by KromekTM, Sedgefield, UK) was utilized. The detector was calibrated in energy and efficiency. Samples of [18F]FDG and [68Ga]-DOTANOC were measured to assess the detector’s suitability in radionuclidic purity measurements. The energy resolution and the efficiency obtained are fully adequate for quality control of positron emission tomography radiopharmaceuticals. Results of [18F]FDG and [68Ga]-DOTANOC samples proved to be comparable to those obtained using high-purity Germanium detectors. The performance of the detector and its very compact size make this type of device an extremely attractive tool for QC applications in radiopharmacy.

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.

The slowing down equation for elastic scattering of neutrons in an infinite homogeneous medium is solved analytically by decomposing the neutron energy spectrum into collision intervals. Since scattering physically smooths energy distributions by redistributing neutron energy uniformly, it is informative to observe how mathematics accommodates the scattering process, which increases entropy through disorder.

The international monitoring system (IMS) has been developed to verify compliance with the comprehensive nuclear test-ban treaty (CTBT) and radionuclide monitoring stations are the smoking gun for underground nuclear explosions identification. One of the main objectives of the verification regime is to improve the sensitivity of measurements providing background reduction and, as a result, decrease the minimum detectable activity (MDA) values. At this ENEA laboratory an anticoincidence system has been developed: the system is comprised of a high purity germanium (HPGe) detector surrounded by a low–background shield and two plastic scintillators able to detect coincident cosmic-ray interactions. Tests conducted with the anticoincidence system have shown a reduction of the Compton continuum that contributes to the spectrum background and the measured MDA values, compared to those obtained using the standard system, decreased for all the four radioxenon isotopes. © 2018, © 2018 Informa UK Limited, trading as Taylor & Francis Group.

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 16.5 MeV 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.

The Monte Carlo code FLUKA is used to simulate the production of a number of positron emitting radionuclides, 18 F, 13 N, 94 Tc, 44 Sc, 68 Ga, 86 Y, 89 Zr, 52 Mn, 61 Cu and 55 Co, on a small medical cyclotron with a proton beam energy of 13 MeV. Experimental data collected at the TR13 cyclotron at TRIUMF agree within a factor of 0.6 ± 0.4 with the directly simulated data, except for the production of 55 Co, where the simulation underestimates the experiment by a factor of 3.4 ± 0.4. The experimental data also agree within a factor of 0.8 ± 0.6 with the convolution of simulated proton fluence and cross sections from literature. Overall, this confirms the applicability of FLUKA to simulate radionuclide production at 13 MeV proton beam energy.

Radiation therapy is a useful and standard tumor treatment strategy. Despite recent advances in delivery of ionizing radiation, survival rates for some cancer patients are still low because of recurrence and radioresistance. This is why many novel approaches have been explored to improve radiotherapy outcome. Some strategies are focused on enhancement of accuracy in ionizing radiation delivery and on the generation of greater radiation beams, for example with a higher dose rate. In the present study we proposed an in vitro research of the biological effects of very high dose rate beam on SK-Mel28 and A375, two radioresistant human melanoma cell lines. The beam was delivered by a pulsed plasma device, a “Mather type” Plasma Focus for medical applications. We hypothesized that this pulsed X-rays generator is significantly more effective to impair melanoma cells survival compared to conventional X-ray tube. Very high dose rate treatments were able to reduce clonogenic efficiency of SK-Mel28 and A375 more than the X-ray tube and to induce a greater, less easy-to-repair DNA double-strand breaks. Very little is known about biological consequences of such dose rate. Our characterization is preliminary but is the first step toward future clinical considerations.

In view of assessing natural radioactivity with on-site quantitative gamma spectrometry, efficiency calibration of NaI(Tl) detectors is investigated. A calibration based on Monte Carlo simulation of detector response is proposed, to render reliable quantitative analysis practicable in field campaigns. The method is developed with reference to contact geometry, in which measurements are taken placing the NaI(Tl) probe directly against the solid source to be analyzed. The Monte Carlo code used for the simulations was MCNP. Experimental verification of the calibration goodness is obtained by comparison with appropriate standards, as reported. On-site measurements yield a quick quantitative assessment of natural radioactivity levels present (40K, 238U and 232Th). On-site gamma spectrometry can prove particularly useful insofar as it provides information on materials from which samples cannot be taken. © 2016 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND