Radiation Measurements最新文献

In the recent past (Andreo et al., 2020), new ionization chamber-specific beam quality correction factors (k_Q,Q0) were published in the literature for the determination of the absorbed dose to water in high energy radiotherapy photon beams. In a previous experimental investigation, made at our laboratory, five various Farmer-type cylindrical ionization chambers, i.e. Exradin A12, IBA FC65-G, NE-2571, PTW 30011 and PTW 30012, were studied to examine their responses with reference to the determination of the absorbed dose to water in an inter-chamber comparison. In the present work, measurements were made to experimentally study the relation or ratio of the k_Q,Q0 correction factors using the five various types of the ionization chambers employed in our previous investigation. In total, nine ionization chambers, two of each type, were included in the present study, with the exception of a single PTW 30011 ionization chamber. The photon beams employed were the 6 MV and the 15 MV beam. The outcomes of this work displayed excellent agreement between the ratio of the ionization chamber-specific k_Q,Q0 correction factors, for the ionization chambers utilized, and the corresponding ratio calculated using the published k_Q,Q0 correction factors. The discrepancy obtained in these comparisons was within 0.1%, for both the photon beams examined, which is much smaller than the uncertainty associated with the determination of this relationship.

Federal Regulations of the United States require that installations possessing sufficient quantities of fissile material to potentially constitute a critical mass, such that the excessive exposure of individuals to radiation from a nuclear accident is possible, shall provide appropriate nuclear accident dosimetry. The American National Standard ANSl/HPS N13.3–2013 Dosimetry for Criticality Accidents provides technical, quality assurance, and performance requirements for nuclear accident dosimeters (NAD). In 2023 the U.S. Navy operated 82 nuclear-powered ships, with the fleet being composed of 11 aircraft carriers, 68 submarines, having a total number of 98 reactors. Since 1968 the U.S. Navy has used fixed nuclear accident dosimeters (FNAD) mounted to the bulkheads surrounding naval nuclear propulsion reactors. Since 1968 the US Navy has used two nuclear accident criticality dosimeters. The first Navy accident dosimeter DT-518/PD was introduced in 1968. It was developed by the Naval Radiological Defense Laboratory in San Francisco, California under leadership of Eugene Tochilin. This dosimeter contains two indium foils for quick dose assessments using shipboard gamma instruments available on nuclear powered vessels and two sulfur pellets/LiF TLD-700 powder for final dose determination at the Naval Dosimetry Center. The newest Navy NAD is the DT-723/PD, which contains indium foil, gold foil, cadmium shielded gold foil, sulfur pellet and a LiF TLD-700 chip. This paper provides a brief description of the measurement procedures, results of the testing of both NADs and comparison of their performance.

Radiophotoluminescent glass dosimeters (RPLDs) are increasingly used, especially to measure the doses in therapeutic beams, but the reported data are mainly related to photon and electron beams, while the data for proton beams are limited and not always consistent.

In this study, dose measurements were performed with two types of RPLDs: GD-352M and GD-302 M (Asahi Techno Glass Corporation). The dose dependence, long-term fading and response in different Spread out Bragg Peak (SOBP) configurations were evaluated. The dose dependence was measured in the middle of the SOBP for doses ranging from 1 Gy to 120 Gy and a linear dose response was shown for both types of RPLDs. The long-term fading of RPLDs irradiated with proton beams determined under normal ambient conditions over a period of one year is negligible. The GD-352M and GD-302M responses were evaluated in different configurations of the SOPB. The fluence-averaged (LET_f) and dose-averaged (LET_d) LET values at the positions of the dosimeters were calculated using the PHITS Monte Carlo code. GD-302 M results showed a small under-response compared to the doses provided by treatment planning system (TPS) (within 1.9 %). GD-352M showed an under-response compared to TPS doses for all investigated ranges and modulations with an average under-response of 20 %, indicating the need for correction factors. Both types showed a slight decrease in the response with an increase in the LET.

The release of radioactive plumes can occur due to nuclear power plant accidents, terrorist attacks, or scenarios in the chemical, biological, radiological, and nuclear defense (CBRN) field. Rescue aircraft may be exposed to ionizing radiation in such contaminated zones. Previous experimental assessments of these scenarios to estimate the potential dangers crew and passengers face are usually intangible. Using predictive software, it is possible to extract information from an external radiation field to which an aircraft would be subject. However, such codes cannot internally estimate the impact of this field on the aircraft. Using simulations based on the Monte Carlo method, it was possible to reproduce the scenario of plumes in the external environment (atmosphere) and, from this scenario, generate relevant information about their impact on the aircraft's internal environment. These assessments comprise the interaction of radiation from a radioactive plume with the aircraft structures and fuel tanks to estimate the radiation doses received by the crew and onboard electronics. This work evaluates the fluence rate, ambient dose equivalent rate (Ḣ*(10)), and effective dose rate (Ė) inside an aircraft flying through a radioactive plume resulting from a nuclear power plant accident. The simulation results suggest that radiation dose rates vary widely depending on the position within the aircraft. The ambient dose equivalent rate for photons varies by approximately 80% depending on the position within the aircraft. These differences reach around 98% for the ambient dose equivalent and effective dose rates for electrons. The data obtained may also be incorporated into risk assessments and support the development of protective measures to counter CBRN events.

One of the major difficulties in using mobile phone screen glass as a retrospective dosimeter is the high intrinsic background signal present in an unirradiated sample. This study aimed to assess the influence of the intrinsic background signal (not erased before readout) on dose estimation using glasses from the touchscreens of Samsung, iPhone, Xiaomi, Huawei, and Realme mobile phones in comparison to readout with prior erasement of the intrinsic background signal (S0). The most critical triage dose of 2 Gy was investigated, and three groups of screens were distinguished based on the value of the S0 signal. Additionally, the dose linearity in the range of 0.5–2 Gy was checked for four models of phones. The results indicate that the estimation of an accidently absorbed dose without prior removal of (S0) from mobile phone screen glass for doses in the range of 0.5–2 Gy with an uncertainty at the level of several percent is possible, even for screens with a high (S0), e.g., the iPhone 6. In an emergency situation, a calibration should be prepared based on data from irradiated samples without removing the intrinsic background signal beforehand in order to minimise the impact of a possible decrease in sensitivity on the results.

Existing nuclear accident dosimeters are based on the activity measurements of the neutron activation products. Those dose measurements require the use of expensive counting equipment. The other significant problem of existing accident dosimeters is that they need to be analyzed in a very short period, due to the short half-life of the activated products. The crystalline L-α-alanine is a well-known material for Electron Paramagnetic Resonance (EPR) dosimetry. It is widely used in radiation laboratories for both reference and transfer dosimetry. Alanine pellets are made from L-α-alanine powder mixed with 10–20% of a binder. The L-α-alanine pellets have ideal dosimetric properties. Alanine is sensitive to alpha, photon, and neutron irradiation. Until recently there was no way to distinguish separate dose contributions from multiple types of radiation. To overcome this shortfall, the multiple alanine pellet holder has been proposed, designed, and tested. It contains cadmium, lead, and ⁶LiF filters. The proposed approach allows for the separation of photon and neutron dose contributions. Use of the developed holder can also reduce negative effects of humidity and direct sun light on the alanine dosimetric properties. The developed holder design can be produced by a 3D printer at a low cost. Here we present and discuss the design and test results of the multiple alanine pellet dosimeter.

Neutron spectrometry plays a crucial role in decommissioning of nuclear sites and in ensuring the safety of workers from radiation exposure. In this study, we associate a proton recoil spectrometer made by a homemade plastic scintillator with triple discrimination capabilities (fast neutrons, thermal neutrons, and gamma rays) with deconvolution methods to obtain accurate measurements of neutron spectra. The purpose of this work is to test and evaluate the performance of three distinct deconvolution codes, namely MAXED (Maximum Entropy Deconvolution), GRAVEL, and MLEM (Maximum Likelihood Expectation Maximization), using two types of input spectra (flat and Watt spectrum). These deconvolution codes are applied to simulated data using the reference MCNP6.2 Monte Carlo code. Comparing the calculated mean squared error (MSE) performed by the three unfolding methods, we find that MLEM seems to perform better than MAXED and GRAVEL. Furthermore, given the calculated MSE values, the unfolded spectrum of Cf-252 is in better agreement with the standard reference spectrum by using Watt spectrum as input spectrum (MSE<1.2 × 10⁻⁶) than using a flat spectrum (MSE<1.3 × 10⁻⁶).

In the Occupational Intakes of Radionuclides (OIR) report series, the International Commission on Radiological Protection (ICRP) published updated biokinetic models and dosimetric data associated with the internal exposure of workers, which are consistent with the recommendations of ICRP Publication 103. The present study focused on the application of the new OIR uranium model published in ICRP Publication 137, which is relevant for the interpretation of measurements of activity in the body and in excreta, and for calculating the committed effective dose. The main objective of this study was to assess the impacts of the new OIR biokinetic models and dose coefficients compared with the old ones based on ICRP Publications 60/78/119. CIEMAT and the University of Salamanca in Spain have been working on re-interpreting in vitro bioassay data obtained from workers with long-term exposure to inhaling uranium oxides during the fabrication of nuclear fuel elements using low enriched uranium, which is now classified as OIR type M/S material. The effects on the retention/excretion models and dose coefficients were studied for ²³⁴U, ²³⁵U, ²³⁸U, and the uranium mixture. Committed effective doses were re-assessed using BIOKMOD code by applying the new OIR excretion model and dose coefficients considering acute and/or chronic inhalation intake scenarios. A reduction by roughly a factor of four was obtained compared with former doses based on ICRP Publications 60/78/119, which is an important effect to take into consideration.

The International Commission on Radiation Units and Measurements (ICRU) has proposed to change the definitions of the operational quantities used for the area and individual monitoring of external exposure in ICRU Report 95. Since introducing new operational quantities into the radiation monitoring may affect the dose assessment using the present personal dosimeters, precise assessments of the influence on monitoring are necessary by characterizing the energy spectrum in the workplace and the energy dependency of the dosimeters to be used. This study measured photon spectra using a NaI(Tl) scintillation detector or a LaBr₃(Ce) scintillation detector at the Japan Research Reactor No. 3 (JRR-3) and Japan Proton Accelerator Research Complex (J-PARC) workplaces at the Japan Atomic Energy Agency (JAEA). The photon energy spectra were obtained by unfolding the pulse height distributions measured at each workplace. The present and new operational quantities were then evaluated and compared using the spectra measured at the workplaces.

In JRR-3, the photon energy spectra were measured at the workplaces in the reactor room and thermal neutron beam hall while the reactor was operated. The ambient dose equivalent rates $\dot{H}$*(10) at the workplaces were 0.2 μSv·h⁻¹ – 13 μSv·h⁻¹. High energy gamma-rays of 6–7 MeV from ¹⁶N were observed near the fuel failure detection system in the reactor room. Those of 7–8 MeV produced by the thermal neutron capture of iron were observed in the reactor room and the beam hall. In J-PARC, the photon energy spectra were measured at the workplaces in the beam tunnel of the 3 GeV synchrotron beam ring five days after the operation. The ambient dose equivalent rates $\dot{H}$*(10) at the workplaces were 35 μSv·h⁻¹ –430 μSv·h⁻¹.

It was found that the new personal doses, H_p, were 10%–20% smaller than the present personal dose equivalents, H_p(10), at the workplaces.

This paper describes the development of 3D-printed dosimeters made from radiochromic materials. Incorporating polymethyl methacrylate (PMMA) as a transparent polymer matrix and leuco-crystal violet as a radiochromic dye, these dosimeters exhibit a color change when exposed to radiation. This change provides a way to visualize and measure three-dimensional dose distributions, which is important for the accurate verification of radiation doses in radiotherapy. The use of 3D printing technology allows these dosimeters to approximate human organ geometries, which may contribute to safer and more effective radiotherapeutic applications. The dosimeters can be customized into various shapes and sizes, are lightweight, and cost-effective, making them appropriate for use in both clinical and research settings in radiotherapy. The fabrication process is detailed, and the dosimeters have been tested at doses up to 100 Gy in X-ray irradiation and analyzed for dose distribution. Bragg-peak measurements from carbon beam irradiation illustrate the dosimeters' capability to detect peak radiation doses with a spatial resolution of approximately 1 mm. These findings indicate the potential of these 3D printed dosimeters to improve the accuracy of radiation delivery, which could positively impact patient outcomes in radiotherapy.