Radiation Measurements最新文献

Microdosimetric techniques are a valuable tool for beam quality monitoring in BNCT, due to their capability to distinguish different contributions to the total dose and provide physics-based quantities related to biological effectiveness of this composite radiation field. To this aim, measurements are generally performed with gas detectors simulating a tissue-equivalent site size between 0.5 and 2 μm. This work presents instead measurements for site sizes up to 10 μm, performed in the thermal neutron field produced by the accelerator-based MUNES source available at INFN-LNL. An avalanche-confinement TEPC with boron doping in the cathode walls was used. Photon and neutron dose fractions were discriminated in the measured dose-weighted distributions based on their different lineal energy range. In the neutron component two separate peaks could be distinguished for site sizes of 5 μm and greater, the origin of which was tentatively related to contributions due to protons and alpha particles. These results allow to assess the impact of increasing site diameter on the measured relative dose contributions and provide valuable reference data for biological modelling and for comparison with solid-state microdosimeters.

This contribution describes the development of a set of numerical methods based on Machine Learning algorithms to generate an automated classification of experimental Thermoluminescence (TL) Glow Curves obtained routinely by Dosimetry Services. This classification will use experimental data historically recorded by Thermoluminescence Dosimeter (TLD) devices and will be based on the search for possible anomalies in the curves. The classifier tool will ease the labelling of experimental data and the detection of anomalies without previous supervision, implying an improvement in the control evaluations in Quality Guarantee Systems often implemented by Dosimetry Services. Furthermore, this study shows that each curve provides information about the status of each dosimeter, and can be used to perform unsupervised classifications of the measurements.

In this paper, Lu_2.97Al_5–xGa_xO₁₂:Ce_0.03 (x = 0, 1, 2, 3) nanophosphors were synthesized using sol-gel method and calcined at 1100 °C for 3 h. The effect of Ga content on the structural, photoluminescence (PL), and notably thermoluminescence (TL) glow curve, dose response, repeatability and fading properties were investigated. X-ray diffraction (XRD) results indicated that all synthesized samples were crystallized in a pure garnet phase. The PL emission spectra exhibited a broad emission band corresponding to the 5 d → 4f (²F_5/2, ²F_7/2) transition of Ce³⁺ ions in the garnet lattice. Furthermore, a significant decrease in emission intensity was observed upon increasing Ga content. The TL characteristics of nanopowders irradiated with β-rays revealed a significant effect of Ga content on the peak position, shape and intensities of TL Glow curves, which can be explained by the reduction of the energy gap and the distribution of trap levels. The dose response linearity in the range of 0.125–100 Gy was examined for different Ga content, revealing a good linear behavior for x = 0 and 1 Ga. Additionally, samples prepared with x = 0, 1, and 2 Ga exhibited a high level of repeatability across a batch of 10 samples. Also, fading studies were performed for 128 h and revealed strong fading in samples synthesized with x = 0 and 1 Ga. These results suggest potential applications of Lu₃Al₅O₁₂:Ce and Lu₃Al₄Ga₁O₁₂:Ce in ionizing radiation dosimetry.

Air kerma length product and air kerma area product are special quantities used in diagnostic radiology. They are measured using special measurement devices – CT-chambers and KAP-meters, in order to calculate quantities related to patient exposure. Appropriate calibration of all measurement devices is of vital importance, and comparisons between calibration laboratories are necessary to prove competence.

It is usually considered adequate to participate in air kerma comparisons to prove capabilities for special quantities, but the literature shows that some problems in calibration procedure can remain unknown. The comparisons directly in terms of special quantities provide additional burden to laboratories, and require special transfer instruments, but they allow checking the whole calibration procedures.

This paper describes a comparison between two calibration laboratories in terms of both air kerma length product and air kerma area product. Both laboratories achieved good results for all radiation qualities, considering the measurement uncertainty. Transfer instruments’ linearity, field size dependence and energy dependence were investigated. Even though the metrological properties of the transfer instruments are worse than the ionization chambers, they can be taken into account by introducing additional measurement uncertainty, performing appropriate corrections or choosing calibration points for the comparison for the values of influence quantities where the transfer instrument response is relatively flat. These comparisons provide additional value to calibration laboratories, but there are still several challenges related to their organization and execution.

Intercomparison exercise is an integral part of the quality assurance programmes in ²²²Rn measurements using active and passive devices. Ensuring the accuracy of data generated through measurements necessitates periodic performance checks. To achieve harmonization of methods, quantify biases and errors in measurements, and identify the reasons for discrepancies, there is a pressing need for periodic intercomparison exercises. To address this need of numerous laboratories, particularly in the Asia-Pacific region, a calibration facility for ²²²Rn measuring devices has been established at the Centre for Advanced Research in Environmental Radioactivity (CARER), Mangalore University, India. An international intercomparison exercise for the ²²²Rn measuring devices was conducted at this facility in the year 2022 with the participation of 10 laboratories from six countries. A total of 354 devices (comprising of both active and passive detectors) were subjected to intercomparison measurements by exposing them to three concentration levels: (5.97 ± 0.29) × 10³ Bq m⁻³, (0.70 ± 0.09) × 10³ Bq m⁻³ and (2.02 ± 0.20) × 10³ Bq m⁻³ in separate experiments and the exposure durations were 7 days, 15 days, and 10 days respectively. The performance of each laboratory was evaluated following the criterion adopted by the International Atomic Energy Agency (IAEA) in its worldwide proficiency test programmes. This exercise brought out the fact that some of the laboratories need to improve their measurement practices to produce reliable data.

OSL fading with storage time after irradiation remains a major obstacle in the development of ideal OSL materials. Dy and Eu co-doped NaMgF₃ are attractive candidates for various rare earth doped matrix materials. In this study, NaMgF₃:Dy,Eu was synthesized by a solid-state reaction, and the effects of heating temperature, duration, atmosphere, and cooling rate on XRD, TL and OSL properties were studied. A simple, safe, efficient, and time-saving solid-state reaction was identified as a potential method for the preparation of NaMgF₃:Dy,Eu, which could be optimally prepared by heating at 750 °C for 2 h under nitrogen atmosphere (2 l/min) followed by rapid cooling. The results show that NaMgF₃:Dy,Eu has a more stable OSL response and an excellent TL glow curve, with only a 0.4% decrease in the OSL signal read after 1 d of dose irradiation compared to that immediately after irradiation, and a high main TL peak at ∼320 °C. It has been indicated the OSL signal in this material seems to be strongly related to the main TL peak. The material has a considerable OSL sensitivity and decay rate than the Luxel Detector. NaMgF₃:Dy,Eu will have a promising future in the field of OSL dosimetry due to its near tissue equivalence, low OSL fading without preheating, fast OSL decay rate, and predictable and easily reusable dose elimination.

The changes in the neutron spectrum produced by the AmBe neutron source moderated in the polyethylene sphere need to be characterized. A passive single-moderator neutron spectrometer with indium foil activation neutron detector has been developed to characterize the neutron spectrum of an AmBe source inside a polyethylene sphere. The detector response function was calculated using the MCNPX 2.7 program with IRDFF-II nuclear data library. Measurements were conducted by placing the Single-Cylindrical Neutron Spectrometer (SCNS) on the outer surface of the sphere for 5 h. The ^116mIn activity due to neutron activation in each indium foil was measured using a gamma spectrometer with an HPGe detector. By identifying the count value at the peak energy of 1294 KeV and considering an HPGe detector efficiency of 3.2% at the foil position 1 mm above the detector surface, the activity of ^116mIn was obtained. The activity value of ^116mIn from each indium foil was compared with the MCNPX simulation results. The neutron spectrum was unfolded using the UMG 3.3 program with activity data input of ^116mIn for each foil and detector response. A neutron spectrum was obtained with a total neutron flux of 634 ± 60 n/cm²·s, consisting of 25% thermal neutrons, 16% epithermal neutrons, and 61% fast neutrons. When compared with the simulation results, the total neutron flux in the spectrum produced by SCNS-In showed only a small difference of 1%. Based on these neutron spectrum measurements, it was determined that placing the AmBe neutron source inside a 15″ diameter PE-sphere will reduce the fast neutron flux by 78%.

A nuclear criticality results in the emission of both neutron and gamma radiation and can produce doses to personnel near the event that exceed 0.1 Gy (10 rad). The primary purpose of nuclear accident dosimetry is to rapidly identify affected personnel in need of prompt medical treatment and to reassure personnel who have been only minimally exposed. While accurate dosimetry is desired, it must be recognized that dose determinations made from whole-body dosimeters or simple triage methods are very rough estimates and contain significant uncertainties. Even when accounting for factors like varying neutron energy spectra, mean photon energies, body orientation within the radiation field, and transient effects on dosimeter response, etc., the end value is a dosimetric quantity defined for very specific radiological conditions and determined within a simple phantom usually at a single depth. Of more importance is the biological response to the radiation, which will vary by person and can be affected by the individual's radiation sensitivity, age, gender, mass, and underlying health conditions. The overall biological, person-specific response to a given dose cannot be precisely determined except by patient symptom observation and individual biological dosimetry (e.g. chromosome analysis, lymphocyte ratios, etc.). This work describes and discusses the criticality accident dosimetry program at the Y-12 National Security Complex, a United States Department of Energy National Nuclear Security Administration facility. The primary goals of the Y-12 accident dosimetry program are, among others, the rapid identification of significantly exposed persons, prompt routing of exposed workers for medical evaluation and treatment, and the ultimate processing of dosimeters to assign doses to personnel.

This study investigates pediatric Computed Tomography (CT) dosimetry, specifically in the thoracic and head regions, using thermoluminescent dosimeters (TLD). The investigation, conducted at the Imaging Diagnostic Center of the Dom Vicente Scherer Hospital, involved two CT machines, the GE Optima 540 and Revolution EVO models. It aims to correlate in vivo TLD analysis with experimental CT Dose Index Volume (CTDIvol) values. The research reveals a maximum variation between machine-indicated CTDI values of 5.7% for the Optima 540 model and 6.8% for the Revolution EVO model at a voltage of 120 kV. Comparison with dose reference levels (DRL) from the United Kingdom and the United States indicates that the obtained values are below these standards, suggesting safe practices in the participating hospital. However, a larger sample size is recommended to establish local standards securely. In the thoracic region, nominal CTDI_vol values indicate measurements around 150% lower than the dose values measured by TLD's, however, no correlation was found between the two variables (p-value = 0.09). In the head, nominal CTDI_vol values varied on average 21% above the doses measured by TDL's, showing a strong correlation between the two quantities (p-value = 0.0002). The study highlights the importance of cautious interpretation of the CTDI_vol and the need for continuous optimization of procedures to ensure safe practices and minimize the risks of radiation exposure in pediatric patients.