An experimental search for a correlation between outdoor 222Rn concentration and 210Pb activity in air particulate samples

This work was aimed to search for a correlation between outdoor 222Rn air concentration and 210Pb activity in particulate samples collected on a filter. The existence of a correlation could support the hypothesis that both 222Rn and its long-lived product 210Pb are embedded into the same air masses and undergo the same air transport phenomena. Lead-210 activity was determined by means of gamma-ray spectrometric measurement carried out a few days from the sampling whereas 222Rn concentration was measured through a commercial monitor. Experimental tests allowed to obtain a weak correlation between 222Rn and 210Pb air concentration as a preliminary result due to high uncertainties of outdoor 222Rn concentration measurements

EVALUATION OF ENVIRONMENTAL NEUTRON DOSE AT GROUND LEVEL

This work presents the results of two cycles of neutron dose rate measurements realized using an ALNOR 2202D Neutron Dose Rate Meter whose time response is acquired and analyzed through a controlled ORTEC MCS-32 acquisition card in Windows environment. The data obtained have been compared with values from previous experimental surveys and with the data provided by the worldwide main observatories. It has been also verified the influence of the fluctuations in the flux of cosmic rays during the course of a solar cycle. By comparing the realized measurements and the data provided by the cosmic ray monitoring networks it is also possible to obtain a value of ambient dose equivalent rate and neutron flux rate which can be used as a reference for design a neutron irradiation testing of electronic devices

Measurement of long-lived radionuclide activity induced in target components of a cyclotron used for [18F ]-[FDG] production

The evaluation of high activity induced in target components in using a medical cyclotron to produce positron-emitting radionuclides for PET (Positron Emission Tomography) diagnostic studies is one of important issue and involves radiation protection concepts when operators are engaged in maintenance and/or substitution of a target or its components. Most of replaced target components are generally classified and stored in a Pb-shielded container in order to wait for their radioactive decay. However, after some years, it can be necessary to start with the removal of the oldest parts, for a temporary storage of the fresh activated ones. The feasibility of these operations and the waste final disposal depends on the level of activity achieved and in particular on the residual concentration of radionuclides. In this work we perform a nuclide identification and activity evaluation of some activated target parts by high resolution gamma-ray spectrometry with various HPGe detectors. The measurements were performed over a decay period of more than 10 years from extraction, which allows to identify radionuclides with different half-lives. In particular, measurements on some Havar foils, stripper foils and titanium parts of a target used inside an IBA CYCLONE 18/9 cyclotron, allow to evaluate the largest activity values related to the most important radionuclides produced by activation of the materials (51Cr, 52Mn, 54Mn, 56Co, 57Co, 58Co, \u2026) with half-life of 70-80 days, while radionuclides with higher half-lives (22Na, 44Ti, 60Co, 207Bi, \u2026..) were detected in the same samples in measurements performed after a long time period. In this way it is possible to evaluate in advance the activity level at a time period after the end of maintenance and establish the correct procedures for storage or disposal of wast

A baseline estimation procedure to improve MDA evaluation in gamma-ray spectrometry

The evaluation of minimum detectable activity (MDA) for a radionuclide in a gamma-ray spectrum is generally carried out through the computation of a suitable background count. This task is sometimes difficult for complex spectra for the presence of many photopeaks which make the trend of continuum extremely variable due to multiple dispersion effects and interference factors. It follows that the MDA assessment must be take into account the contributions of all gamma emissions of radionuclides contained in a sample and its value can be significantly higher than that determined by considering only the background of the spectrometric system due to the overlapping of other peaks. A procedure or an algorithm to determine, each time, the count values to be used for the calculation of MDA is interesting and useful. In this work, some of the more recent algorithms proposed for background subtraction in a gamma-ray spectrum have been examined, applying them in an inverse way for the evaluation of baseline trend in the whole energy range. Among the algorithms examined, particular attention was paid to the application of SNIP (statistical sensitive nonlinear iterative peak clipping) algorithms, which are the simplest to adopt and implement in an application procedure. The results obtained in the analysis of test gamma-ray spectra are satisfactory and allow to quickly determine the MDA values with a formulation based on the ISO-11929 standard

Impact of IFN lambda 3/4 single nucleotide polymorphisms on the cytomegalovirus reactivation in autologous stem cell transplant patients

Cytomegalovirus (CMV) infection represents one of the main cause mortality after Stem Cell Transplantation. Recently, a protective effect of the T allele of rs12979860 IL28B Single Nucleotide Polymorphisms (SNPs) against CMV infection in the allogenic stem cell transplantation was suggested. We investigate whether the rs12979860 IL28B SNP and the relative rs368234815 (IFNλ4) genotype may affect the incidence of active CMV infection in Autologous stem cell transplantation (Auto-SCT) setting. The study included 99 patients who underwent to Auto-SCT. IL28 and IFNΔ4 SNPs were correlated with CMV reactivation along with other clinical and treatment parameters. CMV reactivation by CMV DNAemia was evaluated once a week until day 100 from Auto-SCT. CMV reactivation was documented in 50% (TT-ΔG/ΔG), 35% (CC-TT/TT) and 29.2% (CT-TT/ΔG) of the patients respectively. No differences in CMV copies number were recorded at reactivation between different IL28/IFNλ4 genotypes. The analysis of patients older than 60 years showed a significantly higher incidence of active CMV infection in the TT-ΔG/ΔG (83%) population with respect to CC-TT/TT (21%) and CT-TT/ΔG (40%) patients. Our data suggest a negative role of TT-ΔG/ΔG genotype in the CMV reactivation in Auto-SCT. The exposure to rituximab and the pre-infusion presence of anti CMV IgG also significantly influenced CMV reactivation

Definizione di mappe georeferenziate di dose da esposizioni per contaminazione da radionuclidi rilasciati nell'ambiente atmosferico in seguito ad un ipotetico evento incidentale

Il sistema di calcolo CALMET-CALPUFF è stato utilizzato per lo studio del trasporto a lunga distanza e la deposizione di diverse specie radioattive a seguito di un ipotetico incidente nella centrale nucleare di Gösgen, situata presso il Canton Soletta, Svizzera. CALPUFF è un modello gaussiano a puff in grado di descrivere la dispersione atmosferica di inquinanti multi-specie su scale che vanno dalla decina alle centinaia di chilometri dalla sorgente. I dati meteorologici sono stati simulati con il codice di calcolo 3D CALMET. I risultati in termini di concentrazioni dell’inquinante radioattivo in aria e deposizione al suolo sono stati elaborati con il software GRADO (GRid_in Air of specific effective DOse), messo a punto dal gruppo di ricerca per la definizione di mappe georeferenziate di dose da esposizione utilizzando i coefficienti di conversione riportati nelle pubblicazioni ICRP. L’analisi dei risultati mostra che il pacchetto software integrato CALMET-CALPUFF-GRADO può risultare uno strumento utile nelle strategie di gestione dell’emergenza in caso di eventi incidentali come quello oggetto di studio

Indoor Fast Neutron Generator for Biophysical and Electronic Applications

This study focuses the attention on an indoor fast neutron generator for biophysical and electronic applications. More specifically, the findings obtained by several simulations with the MCNP Monte Carlo code, necessary for the realization of a shield for indoor measurements, are presented. Furthermore, an evaluation of the neutron spectrum modification caused by the shielding is reported. Fast neutron generators are a valid and interesting available source of neutrons, increasingly employed in a wide range of research fields, such as science and engineering. The employed portable pulsed neutron source is a MP320 Thermo Scientific neutron generator, able to generate 2.5 MeV neutrons with a neutron yield of 2.0 x 106 n/s, a pulse rate of 250 Hz to 20 KHz and a duty factor varying from 5% to 100%. The neutron generator, based on Deuterium-Deuterium nuclear fusion reactions, is employed in conjunction with a solid-state photon detector, made of n-type high-purity germanium (PINS-GMX by ORTEC) and it is mainly addressed to biophysical and electronic studies. The present study showed a proposal for the realization of a shield necessary for indoor applications for MP320 neutron generator, with a particular analysis of the transport of neutrons simulated with Monte Carlo code and described the two main lines of research in which the source will be used

Dosimetry to Electron Spin Resonance (ESR) using organic compounds (alanine and ammonium tartrate) for mixed neutron-gamma fields

Alongside with the development of Neutron Capture Therapy (NCT) and the use of thermal neutrons for radiotherapeutic purposes, many efforts have been devoted to the characterization of the beam in order to optimize therapy procedures. Reliable dose measurements should be able to determine the various (neutrons and photonic) components of the mixed beam usually employed for therapy. This paper studies the effect of additives such as Boric and Gadolinium nuclei on the sensitivity of neutron organic (alanine and ammonium tartrate) dosimeters analyzed through Electron Spin Resonance (ESR) technique. These dosimeters were exposed to a mixed (neutron-gamma) field mainly composed of thermal neutrons. The choice of 10B and 64Gd as nuclei additives is due to their very high capture cross section for thermal neutrons. Also, after the nuclear reaction with thermal neutrons are emitted particles, which in turn release their energy in the vicinity of the reaction site. The irradiation with mixed (neutron-gamma) field were performed within the thermal column of the TRIGA reactor, University of Pavia. Dosimeters readout was performed through the Electron Spin Resonance (ESR) spectrometer Bruker ECS106 located at the Laboratory of Dosimetry ESR / TL of the Department of Physics and Chemistry - University of Palermo. We found that the addition of Gadolinium allows to largely increase the sensitivity of the dosimeters for thermal neutrons. In particular, a low concentration (5% by weight) of gadolinium oxide leads to an improvement of the sensitivity of neutrons more than 10 times. In addition, for this low content of gadolinium the photon tissue equivalence is not heavily reduced. This experimental analyses are compared with computational analyses carried out by means of Monte Carlo simulations performed with the MCNP (Monte Carlo N-Particle) transport code. A good agreement was observed for alanine dosimeters