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At high-energy accelerator facilities like the ones that are part of the accelerator complex at the European Organization for Nuclear Research (CERN), Monte Carlo radiation transport codes are widely employed to face the challenges of estimating radionuclide production yields and activities with the aim of performing the radiological characterization of activated components. Indeed, it is of paramount importance to ensure adequate radiation protection during scheduled maintenance, transport, and handling of these components and to establish their proper disposal pathway once they ultimately reach the end of their useful life. This paper summarizes the principles of the fluence conversion coefficients method that was developed as a complementary approach for radiological characterization studies. Then, the Monte Carlo simulations in preparation to the pilot beam run at the Large Hadron Collider at CERN in 2021 are presented as a practical example of possible applications. Finally, the flexibility of the method and the most relevant operational radiation protection implications are discussed in relation to the provided example. [ABSTRACT FROM AUTHOR]

In recent decades, the exposure of aviation crews and passengers to cosmic radiation has been progressively increasing due to the growing need of flights. Therefore, there is a need for radiation assessment software tools to act as a complement to other radiation protection techniques and countermeasures. In this work, the exposure to cosmic radiation is estimated for flights from Athens International Airport to various international destinations, by performing Monte Carlo simulations with the validated tool DYASTIMA/DYASTIMA-R. The results of the estimated total ambient dose equivalent, as well as the ambient dose equivalent rate for different flights, applying a typical flying level and constant atmospheric conditions, are presented for the first time. This study is carried out for different phases of solar activity for the time period from 1996 to 2019, which includes the two recent Solar Cycles 23 and 24. [ABSTRACT FROM AUTHOR]

The radiological protection has the purpose of safeguarding the physical well-being of the user, preventing exposure to detrimental levels of ionizing radiation. This study introduces a novel, cost-effective category of lead-free elastomeric material designed for radiation shielding. The filler compounds utilized are notably lighter than conventional lead-based materials, enhancing user ergonomics during application. They comprise of a blend of barium sulfate combined or not with magnesium oxide with addition-cure liquid silicone rubber. To ensure the effectiveness of the radiation shielding, X-ray transmission measurements were performed for the different thicknesses of the materials and the results compared with Monte Carlo simulations. Additionally, the physical properties of the new materials, such as density, homogeneity, tensile strength, viscosity, and wettability, were also evaluated. The findings indicate that both materials fulfill the requirement for application in radiation protection garments. [ABSTRACT FROM AUTHOR]

Radiation safety and protection is a subject of major concern for all communities because of the biological damage caused by different radiation sources. Therefore, the purpose of this study is aimed to synthesize two different thin films from a newly synthesized N-Al2O3@Pb(II)-2,4-DiNA nanocomposite which was prepared by the direct chemical coupling of nano-aluminum oxide (N-Al2O3) with Pb(II)-2,4-dinitroaniline complex. The synthesized materials were characterized by FT-IR, thermogravimetric analysis (TGA), scanning electron microscope SEM, and elemental composition by energy-dispersive X-ray (EDX). The two thin films were assembled by the encapsulation of N-Al2O3@Pb(II)-2,4-DiNAnanocompositeinto hydrolyzed polyurethane and hydrolyzed polyvinyl chloride. Electron spin resonance (ESR) measurements have been used to investigate the newly prepared nanocomposite and thin films before and after gamma irradiation. Gamma spectroscopy, Phy-X software, and SRIM (stopping and range of ions in matter) Monte Carlo simulation code are aimed to apply and investigate radiation shielding efficiency of films for gamma/x-ray, protons, alpha, and thermal neutrons. The gamma shielding characteristics, mass stopping power (MSP), and range for both proton (H-ions) and alpha (He-ions) were also calculated. In addition, the SRIM code's subroutine TRIM was used to predict the damage production and atomic displacements per atom (dpa) caused by the interaction of high-energy H-ions (proton) and He-ions (alpha) with the studied samples. The data obtained the good radiation shielding effect of the prepared thin films and their possible use as radiation shielding materials for X-rays and low gamma energy. [ABSTRACT FROM AUTHOR]

Background: This study aimed to evaluate the capability of a novel non-Lead composite shield with a lower density compared to commercial Lead shield to protect against 10-100 keV X-ray. Materials and Methods: In the present study, Monte Carlo calculations were performed to study the attenuation properties of a composite designed shield. For this reason, the attenuation rate of 8-material alloy including Tungsten, Bismuth, Barium, Gadolinium, Cesium, Cerium, Tin, and Antimony elements was compared to the 8-layers combination of these elements. Tungsten, Gadolinium, Cerium, and Bismuth were selected to simulate 4-layer and an alloy shields. The resulting X-ray fluence spectra attenuated by the designed shields were calculated. Results: Preliminary results showed that at the energy of 68 keV, the attenuation value by the 8-layer composition was 2.6 times higher than the value achievable by the alloy. It should be noted that the attenuation of 4-layer shield combined from Tungsten, Gadolinium, Cerium and Bismuth in the energy range of 72-100 keV, was 3.5 times higher than that of the alloy shield built from these elements. Conclusions: According to the results, a combination of Tungsten, Gadolinium, Cerium, and Bismuth materials can be effectively used to create non-Lead shields. The four-layer shield with higher density offers better attenuation compared to the alloy shield. [ABSTRACT FROM AUTHOR]

Ionizing radiation is becoming increasingly prevalent in various applications on Earth and in space. This study evaluates the radiation shielding performance of five quaternary glass systems with different chemical compositions of 60SiO 2 -35Pb 3 O 4 -(5-x)ZnO-xWO 3 (0 ≤ x ≤ 5 mol%) using theoretical software and Monte Carlo simulations. The study calculates various shielding parameters, including attenuation coefficients, mean free path, half value layer, tenth value layer, effective atomic number, electron density, and build-up factors, for the photon energy range of 1 keV to 100 GeV. PZSW3 and PZSW5 show comparable attenuation coefficients at low and high energy levels, which are higher than the other selected samples. Additionally, PZSW5 has a neutron removal cross-section of 0.120 cm−1, slightly less than PZSW4's 0.122 cm−1 and greater than the other three samples. On average, PZSW5 (60SiO 2 -35Pb 3 O 4 -5WO 3) exhibits greater radiation attenuation than other samples and most of the compared 19 samples in terms of shielding parameters. The study also analyzes the trajectories of 10 5 photons in the system and finds that PZSW5 has the most effective radiation protection capability among the glass composites studied. [ABSTRACT FROM AUTHOR]

The study aimed to present an alternative method that is applicable to develop the perfect tissue-equivalent material that is completely tissue equivalent in all of its gamma-ray interaction properties. In this work, a Monte Carlo code has been written and simulations have been performed to achieve the gamma-ray interaction data details for the selected body tissues and tissue substitutes quoted from the ICRU Reports that are required for accurate assessment in radiotherapy, radiodiagnosis patient dosimetry and radiation protection. Accurate values of the linear attenuation coefficient for the selected absorber samples have been calculated by using simulation outputs. A comparison has been made with theoretical and experimental data that has appeared in the literature. The results presented in the tables, appear to be close to each other, which shows that the simulation procedure used in this work is suitable for use in such kinds of studies. The results reported here imply that the method presented in this study is important and reliable in providing convenient substitutes for certain treatment procedures for phantoms, radiation detectors, detector walls and tissue-equivalent proportional counters. [ABSTRACT FROM AUTHOR]

The role of dysprosium oxide (Dy2O3)-doped PbO–B2O3–SiO2 glasses on gamma and neutron radiation shielding properties was investigated and evaluated in terms of potential use in radiation protection. The glass composition was arranged according to the formula (55–x)B2O3–25PbO–20SiO2–xDy2O3 (0 ≤ x ≤ 5) (mol%). The Monte Carlo simulation (MCNPX) code was utilised to simulate the mass attenuation coefficient (MAC) of the glasses. When photon energies were in the range of 0.015–15 MeV, the resulting simulated mass attenuation (µm) values were evaluated with conventional Phy-X/PSD database findings. The Phy-X/PSD collected data and the predicted µm findings were in good agreement. The half-value layer (HVL), effective atomic number (Zeff), exposure and energy absorption build-up factors (EBF and EABF) of the glass samples were evaluated. 5 mol% Dy2O3-doped sample has significant γ-ray shielding efficacy in the complete energy range due to lower values of HVL, higher values of MAC and effective atomic number. The build-up factors follow a similar pattern with incident photon energy and penetration depth, growing on a steeply increasing trend up to a maximum at 0.5–1.5 MeV energy, then gradually reducing as photon energy increases up to 15 MeV. The build-up factors show the same routine with penetration depth and incident photon energy. They tend to steeply rise to a peak level at 0.5–1.5 MeV energy, then gradually decrease as the photon energy increases to 15 MeV. Also, the neutron shielding ability was detected and the effective removal cross-section Σ R (cm−1) was estimated theoretically. Due to the greatest value of effective removal cross-section, 5 mol% Dy2O3-doped glass exhibits stronger neutron shielding characteristics than the other glass samples. [ABSTRACT FROM AUTHOR]

Radiation is important in dental science as it is used in both diagnostic and also in treatment issues. The teeth are every single day exposed to a significant amount of wear due to remaining hard particles from food and tooth brushing or even due to the contact with the other teeth. Dental restorations are widely used to correct the damage caused by this wear. In this study, it is aimed to determine the radiation shielding properties of the dosimetric materials which are namely hydroxyapatite (HAP) and urethane di(meth)acrylates (UDMAs). The linear attenuation coefficients of dosimetric materials were calculated using GAMOS (Geant4-based Architecture for Medicine-Oriented Simulations) Monte Carlo simulation code. The obtained results were also compared with the calculation done using XCOM. The mean free path (mfp), half value length (HVL), and tenth value length (TVL) have also been obtained. The highest LAC value was obtained for S2 sample for 511, 662, 835, 1173, 1275, and 1332 keV energies (0.170, 0.149, 0.133, 0.112, 0.108, and 0.105 cm−1, respectively) and the lowest LAC values were obtained for S3 sample for 511, 662, 835, 1173, 1275, and 1332 keV gamma energies (0.102, 0.091, 0.082, 0.069, 0.066, and 0.065 cm−1, respectively). [ABSTRACT FROM AUTHOR]

Individual monitoring of radiation workers is essential to ensure compliance with legal dose limits and to ensure that doses are As Low As Reasonably Achievable. However, large uncertainties still exist in personal dosimetry and there are issues with compliance and incorrect wearing of dosimeters. The objective of the PODIUM (Personal Online Dosimetry Using Computational Methods) project was to improve personal dosimetry by an innovative approach: the development of an online dosimetry application based on computer simulations without the use of physical dosimeters. Occupational doses were calculated based on the use of camera tracking devices, flexible individualised phantoms and data from the radiation source. When combined with fast Monte Carlo simulation codes, the aim was to perform personal dosimetry in real-time. A key component of the PODIUM project was to assess and validate the methodology in interventional radiology workplaces where improvements in dosimetry are needed. This paper describes the feasibility of implementing the PODIUM approach in a clinical setting. Validation was carried out using dosimeters worn by Vascular Surgeons and Interventional Cardiologists during patient procedures at a hospital in Ireland. Our preliminary results from this feasibility study show acceptable differences of the order of 40% between calculated and measured staff doses, in terms of the personal dose equivalent quantity Hp(10), however there is a greater deviation for more complex cases and improvements are needed. The challenges of using the system in busy interventional rooms have informed the future needs and applicability of PODIUM. The availability of an online personal dosimetry application has the potential to overcome problems that arise from the use of current dosimeters. In addition, it should increase awareness of radiation protection among staff. Some limitations remain and a second phase of development would be required to bring the PODIUM method into operation in a hospital setting. However, an early prototype system has been tested in a clinical setting and the results from this two-year proof-of-concept PODIUM project are very promising for future development. [ABSTRACT FROM AUTHOR]

High energy FRagment Separator (HFRS), an essential part of High Intensity heavy-ion Accelerator Facility (HIAF), is used to produce, separate and identify the Radioactive Ions Beams (RIBs). The target area in the pre-separator of HFRS can produce very strong secondary radiation, which seriously threatens the radiation safety of the environment, public and personnel. In this work, the FLUKA code is used to complete the radiation shielding calculation and design of the target area. Firstly, the secondary radiation produced by 25 different typical beam–target combinations are compared. The parameter of 78Kr in fast extraction mode is determined as the basis for the radiation shielding design based on its strongest secondary radiation. Then, the secondary neutron spectrum and the induced radioactivity of the production target and the beam dump are analyzed. The results show that the maximum neutron energy is close to 3 GeV, and special attention should be paid to high-energy neutrons in shielding design, as well as the maximum residual dose rates on the surface of the production target and the beam dump are 3.93 × 107 μ Sv/h and 9.58 × 107 μ Sv/h, respectively, after one month of continuous irradiation. Furthermore, the radiation shielding design is completed using the Q235 steel compact shielding together with the ordinary concrete and sand soil, based on the Monte Carlo simulations and the relevant dose limits. Finally, the radiation effects under this shielding scheme are analyzed. The results show that the shielding design meets the relevant design goals and can provide reliable radiation protection for the environment, public and personnel. [ABSTRACT FROM AUTHOR]

The search for shielding materials against ionizing radiation is very important due to the harmful effect that these radiations cause on human health and the environment. In this work, new ceramic composites of ErBa 2 Cu 3 O x /lead oxide (ErBaCuO/PbO) instead of pure lead are proposed and prepared to serve as promising alternative materials for gamma radiation protection. X-ray diffraction was used to examine the structural properties of the prepared ceramic composites. The analysis showed that all composites have an orthorhombic structure with Pmmm symmetry. The addition of PbO induced distortion of the crystal structure of the ErBCO system. The structural parameters including lattice constants and unit cell volume showed a great variability with the addition of PbO. The physical properties including the density and porosity of the prepared ceramic composites were also determined. It was found that the density values increased from 4510 kg/m3 for 0 wt% to 4640, 4750, 4800 kg/m3 for 2, 5, and 10 wt%, respectively, while the porosity reduced with increasing the PbO content. The Monte Carlo simulation was used to estimate the linear attenuation coefficients of the as-prepared ceramics at different energies of gamma-ray photons ranging from 0.059 to 1.408 MeV. Other shielding parameters were computed, including half-value thickness, transmission factor, and radiation protection efficiency. The results showed an improvement in radiation protection efficiency and a decrease in the transmission factor values with increasing the concentration of PbO. For example, the half-value thickness reduced from 1.949 to 1.832 cm while the radiation protection efficiency increased from 29.93 to 31.51%, at gamma-ray energy of 0.662 MeV, once the concentration of PbO was increased from 0 to 10 wt %, respectively. [ABSTRACT FROM AUTHOR]

Background: In 2016, the International Commission on Radiological Protection (ICRP) published the results of Monte Carlo simulations performed using updated and anatomically realistic voxelized phantoms. The resulting specific absorbed fractions are based on more realistic human anatomy than those computed in the stylized, geometrical Cristy–Eckerman (CE) phantom. Despite this development, the ICRP-absorbed fractions have not been widely adopted for radiopharmaceutical dosimetry. To help make the transition, we have established a correspondence between source and target tissues defined in the CE phantom and those defined in the ICRP phantoms. Results: The ICRP phantom has 79 source regions and 43 target regions in comparison with the 23 source and 18 target tissue regions defined in the CE phantom. The ICRP phantom provides tissue regions with greater anatomical detail. Some of this additional detail is focused on radiation protection and dosimetry of inhaled/ingested radioactivity. Some, but not all, of this detail is useful and appropriate for radiopharmaceutical therapy. We have established the correspondence between CE and ICRP phantom source and target regions and attempted to highlight the ICRP source tissues relevant to radiopharmaceutical therapy (RPT). This paper provides tables and figures highlighting the correspondences established. Conclusion: The results provide assistance in transitioning from CE-stylized phantoms to the anatomically accurate voxelized ICRP phantoms. It provides specific guidance for porting the total absorbed activity for regions as defined in the CE phantom to regions within the ICRP phantoms. [ABSTRACT FROM AUTHOR]

Three zinc barium-borate glasses were produced in the current work, incorporating different metal oxides (MO) (CaO, TiO 2 , and CuO) via the melt quenching technique under a 1100 °C melting temperature for 60 min. The fabricated samples' density is 4.112 g/cm3, 4.157 g/cm3, and 4.261 g/cm3 for glasses dopped with CaO, TiO 2 , and CuO, respectively. Additionally, the Makishima-Mackenzie model was employed to evaluate the effect of metal oxides on the fabricated glasses' mechanical properties. The study shows that the glasses with TiO 2 have the highest mechanical moduli and microhardness which reach 4.67 GPa. The gamma-ray shielding properties in the 0.033–2.506 MeV energy interval were estimated by applying Monte Carlo simulation. The study demonstrates that the glass samples containing CuO exhibit the highest linear attenuation coefficient (LAC) when compared to the glasses containing Ca and TiO 2 compounds. The LAC of the CuO-doped glasses varied between 28.367 cm−1 and 0.163 cm−1 when the photon energy was raised between 0.033 MeV and 2.506 MeV. Also, the study shows good shielding parameters (half-value layer, lead's equivalent thickness, and radiation protection efficiency) for the glasses doped CuO compared to glasses doped Ca and TiO 2. • Zinc barium-borate glasses were prepared for radiation shielding applications. • The gamma-ray shielding properties in the 0.033–2.506 MeV energy range were estimated. • The glass containing CuO exhibits the highest LAC when compared to other glasses. [ABSTRACT FROM AUTHOR]

The hydrothermal technique was utilized to produce a new zirconia doped Bi ions to be suitable for gamma ray shielding applications. The monoclinic phase of zirconia in the manufactured samples is revealed by X-ray diffraction, and each sample's variation in the intensity and position of the diffraction peaks prominently represented the different percentage of BiZrO. Furthermore, the scanning electron microscope and energy dispersive X-ray analysis show the morphology and chemical composition of the fabricated composites. Additionally, utilizing the Monte Carlo simulation, the radiation shielding properties of the fabricated composites were estimated. Regarding the fabricated composites, the highest linear attenuation coefficient of the fabricated composites varied from 21.796 cm−1 to 0.333 cm−1, raising the applied γ-photon energy between 0.059 and 1.332 MeV. Moreover, Addition of Bi 2 to zirconia by concentration varied between 14 and 27 wt% slightly enhances the linear attenuation coefficient from 0.503 cm−1 to 0.539 cm−1, respectively, at γ-photon energy of 0.662 MeV. The slight enhancement in the linear attenuation coefficient increases the radiation protection efficiency and decreases the half value thickness from 1.379 cm to 1.285 cm. [ABSTRACT FROM AUTHOR]

Radon presents significant health risks due to its short-lived progeny. The evaluation of the equivalent lung dose coefficient is crucial for assessing the potential health effects of radon exposure. This review focuses on the uncertainty analysis of the parameters associated with the calculation of the equivalent lung dose coefficient attributed to radon inhalation in mines. This analysis is complex due to various factors, such as geological conditions, ventilation rates, and occupational practices. The literature review systematically examines the sources of radon and its health effects among underground miners. It also discusses the human respiratory tract model used to calculate the equivalent lung dose coefficient and the associated parameters leading to uncertainties in the calculated lung dose. Additionally, the review covers the different methodologies employed for uncertainty quantification and their implications on dose assessment. The text discusses challenges and limitations in current research practices and provides recommendations for future studies. Accurate risk assessment and effective safety measures in mining environments require understanding and mitigating parameter uncertainties. • Inhalation of radon in enclosed spaces poses a risk of lung cancer. • The equivalent lung dose coefficient due to exposure of radon progeny is calculated using the ICRP HRTM. • Quantification of uncertainties on ICRP dose coefficients is important in radiation protection. • Exposure to radon progeny in mines is given in terms of the potential alpha energy concentration. • Uncertainty and sensitivity studies for intake of radon progeny is accomplished using Monte Carlo simulations. [ABSTRACT FROM AUTHOR]

The research aims to exploring the gamma-ray shielding capacities of polyacrylonitrile/chrome-filled polymer composites through a combination of experimental, theoretical and simulation methods. Additionally, employing MCNPv6 and GEANT4 simulation tools, the study evaluates the materials' performance against neutron radiation. The materials were subjected to various gamma-ray energy levels, and their shielding efficacies are analytically quantified using parameters such as Radiation Protection Efficiency (RPE), Mass Attenuation Coefficient (MAC), Linear Attenuation Coefficient (LAC), and Half-Value Layer (HVL). At various neutron energies and sample thicknesses, the numbers of transmitted neutrons were evaluated. Notably, composite P0Cr50 (not contain polyacrylonitrile and containing 50% chromium) emerges prominently, demonstrating superior radiation shielding characteristics against both gamma and neutron radiations. This attitude is attributed to its optimal chrome dispersion and density, positioning it as a promising candidate for radiation shielding applications in industrial and nuclear domains. • Composites series including Polymer/Polyacrylonitrile/Chromium were produced. • Gamma and neutron shielding capacities were evaluated. • Experimental, theoretical and simulation methods were used. • An improvement in radiation shielding was observed with increasing Cr amount. • The composite coded as P0Cr50 has the best radiation shielding. [ABSTRACT FROM AUTHOR]

Composite materials such as metal glasses or alloys combine the useful properties and compensate for the shortcomings of the constituent materials. Therefore, a composite material has improved attributes relative to its individual constituents and therefore, more functionalities. This makes alloys very attractive for radiation shielding in many extreme radiation environments. In this report, different gamma-photon interaction parameters of four SnX alloys (where, X = As, Bi, P, and Sb) were evaluated and analyzed at 0.2, 0.662, and 1.25 MeV. The mass attenuation coefficients (μ ⁄ ρ) of the materials were calculated using the XCOM. Furthermore, other parameters such as the effective atomic number (Z e f f ), electron density (N e f f ), mass energy absorption, the specific gamma constant, absorbed gamma dose rate, and buildup factors were also estimated. The values of μ ρ decreased with energy from 0.2844 cm2/g to 0.0518 cm2/g for SnP, 0.2664 cm2/g to 0.0510 cm2/g for SnAs, 0.3321 cm2/g to 0.0509 cm2/g for SnSb, and 0.7771 cm2/g to 0.0664 cm2/g for SnBi. Also, (Z e f f) S n X and (N e f f) S n X decreased with photon energy within the ranges of 33.01–41.25 and 2.66–3.32 1023 electrons/g for SnP; 41.68–44.30 and 2.59–2.76 1023 electrons/g for SnAs; 56.50–56.51 and 2.329–2.53 1023 electrons/g for SnSb; and 68.26–75.43 and 2.51–2.77 1023 electrons/g for SnBi, respectively. The exposure buildup factor of SnP, SnAs, SnSb, and SnBi at a depth of 40 mfp for 1,25 MeV photons were 27, 30, 17, and 16, respectively. Based on the computed quantities, it is clear that SnBi interact more with photons compared to the other three SnX materials. Moreover, the SnX materials can perform better than some commercial and well-known glass shields, especially at low photon energies. Being Pb-free, the SnX shields are preferable to commercial glass shields containing Pb for radiation protection purposes. • Gamma attenuation of SnX (X = As, Bi, P, and Sb) monolayers. • Radiation shielding performance is studied. • Monte Carlo simulations are employed. • SnX shields are preferable to commercial glass shields. • SnBi monolayer has a potential use in nuclear and medical applications. [ABSTRACT FROM AUTHOR]

In this study, the shielding properties of (Y2O3)0.31−z(Al2O3)z(P2O5)0.69, where 0 ≤ z ≤ 0.31 , in glasses, have been examined to be applied in radiation protection applications. Therefore, we used the Monte Carlo simulation Geant4 code to simulate the mass attenuation coefficient for A, B, C, D, E, F glass samples prepared according to the formula (Y2O3)0.31−z(Al2O3)z(P2O5)0.69. The obtained simulated mass attenuation (µm) results were compared with those obtained theoretically by the XCOM database within photon energies 0.03–2 MeV. The simulated µm results were found to be in good agreement with the XCOM received data. The linear attenuation coefficients(µ), the effective atomic number (Zeff), the effective electron density (Neff), half-value layer (Δ0.5), and mean free path (MFP) were calculated based on the simulated µm. The µm was decreased with the increase of photon energy, and with the increase of (mol%) fraction of Y2O3, and the decrease of (mol%) fraction of Al2O3. It is found that the F glass with 0.364(wt.%)concentration of Y2O3 has superior gamma-ray shielding effectiveness due to its higher values of mass attenuation coefficient, effective atomic number, and electron densities, and lower values of Δ0.5 and MFP. Also, in this work, the neutron shielding ability of the selected glasses was determined. The effective removal cross-section Σ R (cm− 1) was calculated by using the Monte Carlo simulation Geant4 code. Finally, it is found that the B Glass with 0.173 (wt.%)concentration of Al2O3has better neutron shielding properties than the other selected glass samples; due to it has the highest value of effective removal cross-section and has the lowest value of HVL at energy 1MeV. [ABSTRACT FROM AUTHOR]

Background: In the assessment of diseases causing skeletal lesions such as multiple myeloma (MM), whole-body low-dose computed tomography (WBLDCT) is a sensitive diagnostic imaging modality, which has the potential to replace the conventional radiographic survey. Purpose: To optimize radiation protection and examine radiation exposure, and effective and organ doses of WBLDCT using different modern dual-source CT (DSCT) devices, and to establish local diagnostic reference levels (DRL). Material and Methods: In this retrospective study, 281 WBLDCT scans of 232 patients performed between January 2017 and April 2020 either on a second- (A) or third-generation (B) DSCT device could be included. Radiation exposure indices and organ and effective doses were calculated using a commercially available automated dose-tracking software based on Monte-Carlo simulation techniques. Results: The radiation exposure indices and effective doses were distributed as follows (median, interquartile range): (A) second-generation DSCT: volume-weighted CT dose index (CTDIvol) 1.78 mGy (1.47–2.17 mGy); dose length product (DLP) 282.8 mGy·cm (224.6–319.4 mGy·cm), effective dose (ED) 1.87 mSv (1.61–2.17 mSv) and (B) third-generation DSCT: CTDIvol 0.56 mGy (0.47–0.67 mGy), DLP 92.0 mGy·cm (73.7–107.6 mGy·cm), ED 0.61 mSv (0.52–0.69 mSv). Radiation exposure indices and effective and organ doses were significantly lower with third-generation DSCT (P < 0.001). Local DRLs could be set for CTDIvol at 0.75 mGy and DLP at 120 mGy·cm. Conclusion: Third-generation DSCT requires significantly lower radiation dose for WBLDCT than second-generation DSCT and has an effective dose below reported doses for radiographic skeletal surveys. To ensure radiation protection, DRLs regarding WBLDCT are required, where our locally determined values may help as benchmarks. [ABSTRACT FROM AUTHOR]

Glass systems such as Tb2O3-La2O3-P2O5 (TLP-system) exhibit both transparency of visible light and remarkable attenuation of ionizing radiation. In this research article, the TLP-system was explored for potential shielding application against nuclear radiation in various technological facilities. Monte Carlo method is applied by using Geant4 packages to assess the radiation shielding performance of the TLP-system. The validation of the simulation outcomes is achieved by the theoretical calculations. The obtained findings showed that the maximum attenuation coefficients (MACs) of the TLP-system are observed at 0.284 MeV, while the minimum MACs were noticed at 1.275 MeV. At 0.284 MeV photon energy, TLP-A glass with 5 mol% of Tb2O3 has the maximum effective atomic number (Zeff) with worth of 18.3229, whereas the TLP-E glass with 25 mol% Tb2O3 has the maximum Zeff with worth of 21.0558. Moreover, the radiation protection efficiency (RPE) values of the TLP-system decrease with photon energy and increment as the Tb2O3 content increases. [ABSTRACT FROM AUTHOR]

The Radiation protection is of paramount importance in the planning of human exploration activities in space. The related risks must be considered with respect to two aspects: devising a proper shielding and providing answers to the requirement of an effective dosimetry evaluation in astronaut's activities. Both aspects have been considered using the Monte Carlo (MC) code MCNP 6.2 as the reference tool. As case study an application devised for the National Aeronautics and Space Administration (NASA) Artemis program has been chosen. The project aims to establish a sustainable human presence on the Moon, envisioning the realization of an outpost that will serve as a steppingstone for space exploration endeavors. A Class III shelter, in situ resource utilization (ISRU) built habitat for the Moon, has been designed through computational methods and topology optimization techniques, and analyzed in terms of radiation shielding performances and the strictly related structural behavior. The outpost must be able to withstand temperature variations, micrometeorite impacts, and the absence of a substantial atmosphere. Any solution studied to respect the constraints must devise robust and innovative materials and techniques to create habitats that have as goal the shielding from the Galactic Cosmic Rays (GCR) and from the solar flares to provide a safe and habitable environment at the time scales scheduled for the missions. Moreover, the outpost design must incorporate strategies for extracting and utilizing local resources. Overcoming such challenges will pave the way for the establishment of a sustainable human presence on the Moon and serve as a crucial leap for future space exploration missions. • Radiation protection's crucial role in space exploration planning. • Integration of radiation protection by design in lunar habitat design. • MCNP 6.2 and topology optimization coupling. • Design challenges for a sustainable lunar outpost in Artemis. • Innovative materials for shielding cosmic rays and solar flares. [ABSTRACT FROM AUTHOR]

We investigate the effect of neodymium and lanthanum on zirconia radiation shielding and structural properties. To achieve this, new La/Nd-doped ZrO 2 nanomaterials were synthesized and analyzed via various experimental techniques, including energy-dispersive X-ray spectroscopy, scanning electron microscopy, transmission electron microscopy, and X-ray diffraction, to determine the fabricated materials' structure, chemical composition, and morphology. Additionally, to enable the estimation of the effect of the ZrO 2 phase transition on the synthesized La/Nd-doped ZrO 2 nanomaterials' γ-ray shielding properties, Monte Carlo simulation was used. The simulated data revealed that the fabricated materials' linear attenuation coefficient decreased slightly from 17.584 cm−1–17.442 cm−1 at 0.059 MeV, from 0.382 cm−1 to 0.375 cm−1 at 0.662 MeV, from 0.253 cm−1 to 0.248 cm−1 at 1.408 MeV, and from 0.180 cm−1 to 0.176 cm−1 at 15 MeV with decrease of the fraction of the tetragonal phase from 84 % to 57 %. The reduction in the linear attenuation coefficient is associated with increased thickness in terms of the fabricated materials' Pb equivalent and half values. • Neodymium–Lanthanum nanocomposite with doped zirconia synthesized. • XRD, SEM, TEM, and EDX characterization of the composites. • Monte Carlo simulation for estimating attenuation capacity. • Significant enhancement in γ-photon shielding properties. [ABSTRACT FROM AUTHOR]

The estimation of induced radioactivity and radiological classification of activated material is necessary throughout the life cycle of the Large Hadron Collider (LHC) accelerator. As part of the upgrade to High-Luminosity LHC (HL-LHC), radiation protection assessments need to be performed in order to estimate the radiological hazard during the dismantling activities of the LHC cables. The proposed In-situ radiological classification and characterization of the activated cables relies on validated FLUKA.CERN Monte Carlo simulations and analytical computations. The objective is to establish a methodology that allows clearing the material from regulatory control according to the Swiss Radiation Protection Legislation using the total gamma counting (TGC) analysis technique for an envelope set of activation scenarios. The study also includes a preliminary benchmark of the FLUKA geometry model and physics models based on gamma spectrometry laboratory analyses of representative samples. • A new methodology to clear cable material from regulatory control in Switzerland. • Total Gamma Counting analysis techniques with the conservative fingerprint. • Hazard factor 3D maps for radiological zoning and cable classification. [ABSTRACT FROM AUTHOR]

The objective of this study was to examine the radiation protection capabilities of glass specimens with the composition xBaO: 1.5Dy 2 O 3 : 25ZnO: 5Al 2 O 3 : 30B 2 O 3 : (x-38.5)SiO 2 by varying x = 10 (Ba10) to 35 (Ba35) mol% of BaO concentrations. The glass specimens were fabricated using the melt-quenching technique. The results revealed that the physical and optical properties, such as density (ρ), molar volume (V m), and refractive index (RI), had a similar trend. The % transmission result of glass specimens indicates higher transparency than those of commercial and x-ray windows. The experimental values of total attenuation (μ m) in this study used the Compton scattering technique. The obtained results were then compared with theoretical calculation obtained from the WinXCom program and the PHITS (Particle and Heavy Ion Transport Code System) Monte Carlo simulation. The results indicated the values of μ m , effective atomic number (Z eff), and electron density (N eff) increase with the rise in BaO content. However, these values decrease at higher energy levels. The observed trends align with the theoretical predictions of the WinXCom program and the obtained results from the PHITS Monte Carlo simulation. The half-value layer (HVL) showed better shielding properties than both standard radiation shielding glass and other materials. Additionally, an examination of energy absorption (EABF) and exposure buildup factors (EBF) revealed a distinct advantage associated with the incorporation of BaO atoms. This advantageous effect was observed in the reduction of photon scattering events, further enhancing the overall efficacy of the shielding properties in the glass specimens under study. • High-transparency borosilicate glasses was developed as nuclear radiation protection material. • 137Cs used for radiation shielding properties investigation via the Compton scattering technique. • Total attenuation experimental values compared to WinXCom and PHITS Monte Carlo simulation. • Replacing of BaO in glass matrix can enhance shielding properties of the glass. [ABSTRACT FROM AUTHOR]

The work studies the possibility of using a polymer composite based on polytetrafluoroethylene and modified radiation shielding fillers for protection against neutron radiation in space conditions. The physico-mechanical properties of the composite are presented, as well as the simulation of the neutron radiation passage through the polymer composite is performed by the Monte Carlo method using the GEANT4 software toolkit. The polymer composite proposed has high structural properties necessary for use in the space sphere. The Vickers hardness for a load of 200 g is 12.11, the erosion wear is 0.049∙10−3 mm3/g, the density is 4232 kg/m3. The most effective thickness of the protective layer to attenuate the intensity of neutron radiation is 8 cm, to reduce the equivalent dose of 30 g/cm2. The thickness of the protective layer made of the proposed polymer composite against cosmic radiation is 35% less than that of aluminum and 77% less than that of polyethylene, which is a very important factor for using the material in spacecrafts because of extremely limited free space on board. This indicates the prospects of its use for neutron radiation shielding at high energies to achieve complete and effective protection of astronauts and spacecraft equipment. • A polymer composite for protection against neutron radiation has been studied. • The neutron impact was simulated using the GEANT4 software toolkit. • Semi-experimental irradiation of the composite with neutrons was carried out. • Composite retains high erosive wear resistance and Vickers hardness at low density. • The use of a composite will reduce the thickness of the protective layer on spacecraft. [ABSTRACT FROM AUTHOR]

In the present work, the role of Neodymium (III) nitrate hexahydrate [Nd (NO 3) 2 ] and lanthanum nitrate hexahydrate [La (NO 3) 3 ·6H 2 O] on the structural and radiation shielding properties of zircon mineral was examined. Therefore, a new La/Nb-doped ZrO 2 nanocomposite was fabricated. Various experimental techniques such as energy-dispersive X-ray, scanning electron microscope, transmission electron microscopy, and X-ray diffraction pattern were utilized to discover the structure and chemical composition, as well as the morphology of the fabricated La/Nb-doped ZrO 2 nanocomposites. Additionally, for the estimation of the fabricated composites' γ-ray shielding properties, Monte Carlo simulation was employed. Among low and intermediate γ-photon energy, the study affirmed a good shielding characteristic for the fabricated composites, where the linear attenuation coefficient reduced from 245.64 to 0.21 cm−1 when the γ-photon energy was raised respectively from 0.022 to 2.506 MeV. Compared to pure lead, the fabricated composites showed good shielding properties, where the shielding capacity reached 30–50% of the pure lead element in the 0.244–2.506 MeV intermediate energy interval. • Novel Nd/La-doped zirconia nanocomposite synthesized. • Characterization through XRD, SEM, TEM, and EDX. • Monte Carlo simulation estimated material's attenuation capacity. • Significant shielding enhancement for γ-photon energy. [ABSTRACT FROM AUTHOR]

The objective of this study was to analyze the dependence of the neutron dose from the geometry of the second band of the maze using dosimetric measurements of neutrons and Monte Carlo simulations, and based on those results to design a novel radiotherapy room layout. Measurements of the neutron dose at a two-band maze therapy room were performed for a 15 MeV photon beam only. Monte Carlo simulations were performed using the GEANT4 toolkit. In order to obtain the geometry dependence, we were changing the second band angle while we kept the length, height, and width the same as in reality. Results show that the highest calculated dose was obtained for the 60° angle of the second maze. It is 17 % higher than for standard 0° angle. For 30° it was 30 % smaller and for 90° was 10% smaller. Although the lowest dose was obtained for 30° band angle with calculations, it is not very practical for clinical use. Clinically the most interesting would be the 90° angle which is practically a short three-band maze, which could be promising from the perspective of neutron radiation protection since it could offer a compact constructional solution, and better optimization of the available space. [ABSTRACT FROM AUTHOR]

The aim of this paper is to examine the energy and angular responses of thermoluminescent (LiF:Mg,Ti and LiF:Mg,Cu,P) and optically stimulated luminescent (Al2O3:C) dosimeters with experimental measurements and Monte Carlo simulations. Nine radiation qualities, with mean energies ranging from 33 keV to 1.25 MeV, and five angles of incidence, between 0° and 80°, were used to conduct this analysis. The IEC 62387:2020 international radiation protection standard was used as the dosimeter response measure of quality. The experimental and simulated data exhibit that the dosimeter responses meet the standard's criteria, with certain exceptions on lower energies. [ABSTRACT FROM AUTHOR]

Dose equivalent limits for single organs are recommended by the ICRP (International Commission for the Radiological Protection publication 103). These limits do not lend themselves to be measured. They are assessed by convoluting conversion factors with particle fluences. The Fluence-to-Dose conversion factors are tabulated in the ICRP literature. They allow assessing the organ dose of interest using numerical simulations. In particular, the literature lacks the knowledge of local skin equivalent dose (LSD) coefficients for neutrons. In this article, we compute such values for neutron energies ranging from 1 meV to 15 MeV. We use FLUKA, MCNP and GEANT4 Radiation transport Monte-Carlo simulation codes to perform the calculations. A comparison between these three codes is performed. These calculated values are important for radiation protection studies and radiotherapy applications. [ABSTRACT FROM AUTHOR]

Evaluation of the mechanical and shielding features is performed for glasses with the composition xBi2O3-(80-x) TeO2-10GeO2-10B2O3 with different concentrations x = 40, 50, 60, and 65 mol%. The mechanical parameters are summarized utilizing the Makishima-Mackenzie model. The elastic moduli such as Young, bulk, shear, and longitudinal are computed. The obtained results depict the elastic moduli affected by the replacement TeO2 with Bi2O3 content. The maximum values of elastic moduli are 61.822, 40.706, 24.790, and 73.761 for Young, bulk, shear, and longitudinal, respectively, of the inspected glasses utilizing the Monte Carlo simulation code (MCNP-5). The mass attenuation coefficient (μρ) and the half-value thickness (HVT) were also calculated. The shielding ability is tested by comparing the mean free path of the studied glasses with the commercial glasses. Likewise, the accumulation of photons inside the glass's material is determined using the BXCOM program. Finally, it can deduce the discussed fabricated glasses are suitable to apply in diverse applications of radiation protection. [ABSTRACT FROM AUTHOR]

Objective: To estimate radiation doses for the primarily irradiated organs/tissues of patients subjected to standard endovascular aneurysm repair (EVAR) procedures using a novel personalized dosimetry method.Methods: Dosimetric and anthropometric data were collected prospectively for eight patients who underwent standard EVAR procedures. Patient-specific Monte Carlo simulations were performed to estimate organ/tissue doses from each of the fluoroscopic and digital subtraction angiography acquisitions involved in EVAR. Individual-specific cumulative absorbed doses were estimated for the skin, spinal bone marrow, heart, kidneys, liver, colon, bladder, pancreas, stomach, and spleen and compared to corresponding values estimated through a commercially available dosimetric software package that employs standardized phantoms.Results: The highest organ/tissue radiation doses from EVAR were found for the skin, spinal bone marrow, kidneys, and spleen as 192.4 mGy, 96.7 mGy, 72.9 mGy, and 33.6 mGy respectively, while the doses to the heart, liver, colon, bladder, pancreas, and stomach were 6.3 mGy, 14.4 mGy, 18.4 mGy, 14.8 mGy, 21.6 mGy, and 11.2 mGy respectively. Corresponding dose values using standardized phantoms were found to differ up to 151%.Conclusion: Considerable radiation doses may be received by primarily exposed organs/tissues during standard EVAR. The specific size/anatomy of the patient and the variation in exposure parameters/beam angulation between different projections commonly involved in EVAR procedures should be taken into account if reliable organ dose data are to be derived.Key Points: • A novel patient-specific dosimetry method was utilized to estimate radiation doses to the primarily irradiated organs/tissues of patients subjected to standard endovascular aneurysm repair procedures. • The use of standardized mathematical anthropomorphic phantoms to derive organ dose from fluoroscopically guided procedures may result in considerable inaccuracies due to differences in the assumed organ position/volume/shape compared to patients. [ABSTRACT FROM AUTHOR]

This work concerns study of self-absorption factor (SAF) and dose rate constants of zirconium-89 (89Zr) for the purpose of radiation protection in positron emission tomography (PET) and to compare them with those of 18F-deoxyglucose (18F-FDG). We analyzed the emitted energy spectra by 18F and 89Zr through anthropomorphic phantom and calculated the absorbed energy using Monte Carlo method. The dose rate constants for both radionuclides were estimated with 2 different fluence-to-effective dose conversion coefficients. Our estimated SAF value of 0.65 for 18F agreed with the recommendation of the American Association of Physicists in Medicine (AAPM). The SAF for 89Zr was in the range of 0.61-0.66 depending on the biodistribution. Using the fluence-to-effective dose conversion coefficients recommended jointly by the American National Standards Institute and the American Nuclear Society (ANSI/ANS), the dose rate at 1 m from the patient for 18F was 0.143 mSv-MBq-1-hr-1, which is consistent with the AAPM recommendation, while that for 89Zr was 0.154 mSv-MBq-1-hr-1. With the conversion coefficients currently recommended by the International Committee on Radiological Protection (ICRP), the dose rate estimates were lowered by 2.8% and 2.6% for 89Zr and 18F, respectively. Also, we observed that the AAPM derived dose is an overestimation near the patient, compared to our simulations, which can be explained by the biodistribution nature and the assumption of the point source. Thus, we proposed new radiation protection factors for 89Zr radionuclide. [ABSTRACT FROM AUTHOR]

This study explored the zinc-boro-vanadate (ZBV) glass reinforced with Ag2O to show radiation shielding competencies. A glass system with composition 20ZnO-55B2O3-(25-x)V2O5-xAg2O, where x ranges from 0 to 10 with the steps of 2.5 mol%, was designed. Some physical and optical calculations were employed to figure out glass density (ρglass), molar volume (Vm), oxygen molar volume (OMV), oxygen packing density (OPD), refractive index (n), dielectric constant (ϵ), and metalization criterion (M) for the glass system of ZBV0 to ZBV10. According to the findings, one can specify that the increasing content of Ag2O enhanced both the physical and optical features of the ZBV series. The glass density was increased from 3.3150 to 3.6930 g/cm3, while the refractive index was raised from 3.0693 to 3.1216 with the insertion ratio of Ag2O from 0 to 10 mol%. The ZBV10 sample having the maximum Ag2O of 10 mol% ensured to improve physical and optical properties. The shielding properties of the examined glass samples were investigated to evaluate their ability for radiation protection purposes. The Monte Carlo N-Particle Transport Code 5 (MCNP-5) was utilized to assess the projected gamma photons' average track length on the investigated glass samples. Then, the linear attenuation coefficient (LAC), half-value layer (HVL), transmission factor (TF), and radiation protection efficiency (RPE) were calculated based on the simulated gamma photons' average track length. The obtained results showed that the highest LAC values increased from 71.999 to 91.979 cm−1. Moreover, the thicker HVL decreased from 9.869 to 9.332 cm−1, increasing the Ag2O insertion ratio between 0 and 10 mol%, respectively. The mentioned data confirmed that substitution of the V2O5 by the Ag2O enhance the gamma-ray shielding capacity of the explored series of glass. [ABSTRACT FROM AUTHOR]

Determining the neutron source strength is essential to ensure optimal use of the source and to implement adequate radiation protection measures. In this study, the source strength of two (α ,n) neutron sources – a newly acquired 185 GBq 241AmBe and a legacy source 74 GBq 239PuBe – were verified. Source strength verification was performed by measuring the induced activities in irradiated 115In foils and comparing the results to MCNP calculated reaction rates. The results for 241AmBe agree well with a maximum and average deviations of 8.1% and 4.1%, respectively. While for 239PuBe the maximum deviation is 11.7% and the average deviation is 6.3%. Results of this study provide a reliable reference for the application of the sources, in particular, for the 241AmBe which is used as the external neutron source of the PRR1-Subcritical Assembly for Training, Education, and Research (PRR1-SATER), and for the design of a neutron irradiator that will repurpose the 239PuBe. • The source strengths of two (α,n) neutron sources, 241AmBe and 239PuBe were verified through foil activation and Monte Carlo simulation. • Calculated reaction rates and measured saturation activities for the 241AmBe neutron source agree well, with an average deviation of 4.1%. • For the legacy 239PuBe neutron source, calculated and measured results have an average deviation of 5.3%. • The study provides a reliable reference for determining neutron source strength to facilitate efficient utilization of the source. [ABSTRACT FROM AUTHOR]

The borotellurite glasses whose chemical structure is (29.5–0.4x)CaO + 10CaF 2 + (60–0.6x)B 2 O 3 + xTeO 2 + 0.5Yb 2 O 3 (where x=10, 16, 22, 31, and 54 % mole. represent TCCBY1-TCCBY5, respectively) are Pb-free, thermally stable, and transparent glasses with attractive optical features for technological applications. The gamma-photons, electrons, protons, neutrons, carbon ions, fast neutrons, and fast neutron interaction parameters of these glasses are presented in this study to better understand the role of TeO 2 in influencing their radiation shielding properties and radiation protection applications. The photon mass attenuation coefficient was evaluated by XCOM computation and simulation using the FLUKA code. The FLUKA code was also used to evaluate the mass stopping powers of the charged radiations, while neutrons' cross sections were evaluated using standard expressions. For 0.015 MeV–15 MeV photons, the mass attenuation coefficients of the glasses fell from 17.9499 to 0.0246 cm2/g for TCCBY1, 20.5628 to 0.0263 cm2/g for TCCBY2, 23.2756 to 0.079 cm2/g for TCCBY3, 26.7487 to 0.0298 cm2/g for TCCBY4, and 33.3591 to 0.0335 cm2/g for TCCBY5. The photon half-value layer at 15 keV is reduced by about 19.57%, 32.68%, 48.84%, and 63.89% when the TeO 2 content increases from 10 mol to 16, 22, 31, and 54 mol, respectively. TeO 2 was found to suppress photon buildup in the glasses. The mass stopping powers of charged radiation increased as glass density decreased. The addition of TeO 2 into the glass structure increased the ability of the TCCBY glass to absorb fast neutrons by up to 54 % mole. The gamma radiation and fast neutron moderating ability of TCCBY5 glass compared to common shields and other materials is exceptional. The glass is recommended for the design of Pb-free, transparent, and efficient radiation protection structures. • Radiation shielding capacity of borotellurite glasses is estimated. • Monte Carlo simulations are carried out. • HVL is reduced by 63.89% at TeO 2 = 54 mol. • TCCBY5 glass system has a potential use in nuclear applications. [ABSTRACT FROM AUTHOR]

Humans are constantly exposed to ionizing radiation and radioactive substances that exist naturally within the earth, their bodies, and their homes. Therefore, people and their surroundings must be protected from high levels of ionizing radiation. In general, different materials for radiation protection are used in diverse fields and for specific cases. In this work, a new ceramic-based material containing barium tungstate-doped with different amounts (x= 0–75%) of holmium oxide Ho 2 O 3 was proposed as a γ-radiation shield. The structural and optical studies were performed based on X-ray diffraction technique, Fourier transform infrared spectrum, and diffuse reflectance spectroscopies. Radiation attenuation coefficients were determined using the Monte Carlo simulation. The structural analyses indicated that the BaWO 4 crystal belongs to the scheelite structure and a lattice distortion was shown due to the effect of Ho 2 O 3 doping. The band gap energy of the material increased progressively with increasing the content of the dopant in the host material. The radiation shielding coefficients reduced with increasing the concentration of Ho 2 O 3 into BaWO 4 , where the linear attenuation coefficient reduced by 13% from 0.386 cm−1 to 0.338 cm−1 at 0.662 MeV when the concentration of Ho 2 O 3 increased from 0% to 75%, respectively. Simultaneously, the lead equivalent thickness increased from 3.22 cm to 3.67 cm with raising the Ho 2 O 3 concentration. [ABSTRACT FROM AUTHOR]

As an excellent tool, neutron generators have been used widely in prompt gamma-ray neutron activation analysis (PGNAA) technology. In this work, a new PGNAA laboratory was proposed to equip a D–T neutron generator. For safety reasons, radiation shielding of the laboratory was designed, and its protection efficiency was validated by using MCNP simulations and experiments. The results show that the proposed shielding can meet the radiation protection set by the International Commission on Radiological Protection and the Chinese government. [ABSTRACT FROM AUTHOR]

In X‐ray imaging, anti‐scatter grids are used to reduce scatter radiation reaching image receptors, hence improving image quality. Optimization of grid performance is essential for improving image diagnostic quality and minimizing radiation doses to patients. This work investigated the performance of a series of grid designs modeled from the design of typically focused grid with grid ratio 8:1 (r8) and strip height 1.7 mm (h1.7) for high‐energy radiographic applications. Monte Carlo simulation was used to evaluate designs (r8h1.7) which had the strip thickness changed from 6 to 150 μm in 2 μm increments and the interspace distance fixed at 214 μm. The transmissions of radiation in grid materials were modeled by using a regression with radial‐basis‐function‐networks (RBFNS). KSNR was then determined from RBFNS models of radiation transmissions. The optimal strip‐thickness was obtained at the maximum signal‐to‐noise ratio (SNR) improvement factor (KSNR). For high‐energy applications at 100 peak‐kilo‐voltage (kVp) and 30 cm PMMA thickness, the optimal lead‐strip‐thickness was found approximately 74 μm resulting in a strip‐frequency approximately 35 per cm (N35). Using the optimal thickness for imaging condition at 100 kVp and 30 cm thickness, the KSNR would increase by approximately 5.3%. This work showed the existence of optimal strip‐thickness for a series of grids with a given grid‐ratio, strip‐height, strip‐, and interspace materials. The findings are useful and provide guidance to improve grid designs for better performance that will essentially lead to better image quality and better radiation protection for patients. [ABSTRACT FROM AUTHOR]

A new radioactive liquid waste cementation facility was under commissioning recently in the Institute of Nuclear and New Energy Technology of Tsinghua University, which is designed to simultaneously process multiple intermediate-level radioactive waste drums. Therefore, the multiple volume sources and the scattering effect becomes a key issue in its radiation protection. For this purpose, the Monte Carlo program FLUKA code and experimental measurement were both adopted. In the FLUKA simulation, five different scenarios were considered, i.e., one drum, two drums, four drums, six drums, and eight drums. For the multiple volume sources, the source subroutine code of FLUKA was rewritten to realize the sampling. The complex shielding also leads to a deep penetration problem; hence, the optimization algorithm and variance reduction techniques were adopted. During the measurement, two scenarios, outdoor and indoor, were carried out separately representing the dose field when only one drum is considered and when the scattering effect is considered. A comparison between the experiments and calculations shows very good agreement. From both of the Monte Carlo simulation and the experimental measurement, it can be drawn that, in the horizontal direction, with the increase of the drum number, the dose rate increases very little, while in the vertical direction, the increase of the dose rate is very obvious with the increase of the drum number. The complicated source term sampling methods, the optimization algorithm and variance reduction techniques, and the experimental verification can provide valuable references for the similar scattering problem in radiation protection and shielding design. [ABSTRACT FROM AUTHOR]

We numerically study the computed tomography dose index (CTDI) quantity based on the Monte Carlo method using GATE software. In this work, it was demonstrated that the CTDI values decreased following an exponential form as a function of phantom diameter. As expected, the absorbed dose is shown to have a good relationship which increases linearly with X-ray tube current (mAs) values. The simulation presented in particularly that the (CTDI) dose increases not-linearly dependence with photon deposited energy (kVp). It seems that the average percent of the absorbed dose in the abdominal phantom was lower than the heat phantom object's absorbed dose, which was equal to 80%. In conclusion, the use of Monte Carlo simulation represents a dosimetry tool for radiation protection in the field of radiology imaging. [ABSTRACT FROM AUTHOR]

Objectives: To investigate the effect of an organ-based tube current modulation (OTCM) technique on organ absorbed dose and assess image quality in pediatric CT examinations. Methods: Four physical anthropomorphic phantoms that represent the average individual as neonate, 1-year-old, 5-year-old, and 10-year-old were used. Standard head and thorax acquisitions were performed with automatic tube current modulation (ATCM) and ATCM+OTCM. Dose calculations were performed by means of Monte Carlo simulations. Radiation dose was measured for superficial and centrally located radiosensitive organs. The angular range of the OTCM exposure window was determined for different tube rotation times (t) by means of a solid-state detector. Image noise was measured as the standard deviation of the Hounsfield unit value in regions of interest drawn at selected anatomical sites. Results: ATCM+OTCM resulted in a reduction of radiation dose to all radiosensitive organs. In head, eye lens dose was reduced by up to 13% in ATCM+OTCM compared with ATCM. In thorax, the corresponding reduction for breast dose was up to 10%. The angular range of the OTCM exposure window decreased with t. For t = 0.4 s, the angular range was limited to 74° in head and 135° for thorax. Image noise was significantly increased in ATCM+OTCM acquisitions across most examined phantoms (p < 0.05). Conclusions: OTCM reduces radiation dose to exposed radiosensitive organs with the eye lens and breast buds exhibiting the highest dose reduction. The OTCM exposure window is narrowed at short t. An increase in noise is inevitable in images located within the OTCM-activated imaged volume. Key Points: • In pediatric CT, organ-based tube current modulation reduces radiation dose to all major primarily exposed radiosensitive organs. • Image noise increases within the organ-based tube current modulation enabled imaged volume. • The angular range of the organ-based tube current modulation low exposure window is reduced with tube rotation time. [ABSTRACT FROM AUTHOR]

64Cu is an emerging theranostic radioisotope which is generating interest in the field of PET imaging and planning of therapy. A prototype solid target is being designed for producing 64Cu at RMC, BARC. This study evaluates radiation safety aspects involved in the use of the new solid target using FLUKA Monte Carlo code. The radiological hazards are quantified during the production of 64Cu and further handling of the target assembly for post irradiation processing. Analysis of ambient dose equivalent rates from the target assembly shows that during the sufficient quantity of 64Cu production, radiation fields outside the cyclotron vault at RMC BARC are below the allowed limit. The studies also confirm that the 4 h decay after the end of bombardment is suitable for post irradiation handling of the target. [ABSTRACT FROM AUTHOR]

Collimating apertures are used in proton therapy to laterally conform treatment fields to the target volume. While this is a standard technique in passive spreading treatment heads, patient-specific apertures can supplement pencil-beam scanning (PBS) techniques to sharpen the lateral dose fall-off. A radiation protection issue is that proton-induced nuclear reactions can lead to the formation of radionuclides in the apertures. In the experiments of the current study, cylindrical, thick brass targets were irradiated with quasi-monoenergetic proton fields of 100.0 MeV and of 226.7 MeV in PBS mode. The radioactivation of these two brass samples was characterised with a low-level gamma-ray spectrometer. The activation products were scored in a Monte Carlo simulation, too, and compared with the experimental activities. For the high-energy field, 63Zn, 60Cu, and 61Cu were the most important short-lived isotopes regarding the measured specific activity. After irradiation with the 100.0 MeV field, 62Cu, 63Zn, and 60Cu had the highest activity. Regarding long-lived isotopes, which determine the storage time of the used apertures, the isotopes 57Co, 65Zn, 54Mn, 56Co had the largest contribution to the activity. The relative difference of activities between simulation and experiment was typically between 10%–20% for short-lived nuclides and were up to a factor of five larger for long-lived nuclides. Summarising experiments and simulations for both incident proton energies, 62Cu was the most important detected residual nucleus regardless if specific activity or equivalent dose is considered. [ABSTRACT FROM AUTHOR]

Purpose: The determination of x‐ray spectra near the maze entrance of linear accelerator (LINAC) rooms is challenging due to the pulsed nature of the LINAC source. Mathematical methods to account for pulse pile‐up have been examined. These methods utilize the highly periodic pulsing structure of the LINAC, differing from the effects of high‐intensity radioactive sources. Methods: Sodium iodide (NaI) and plastic scintillation detectors were used to determine the energy spectra at different points near the maze entrance of a medical LINAC. Monte Carlo calculations of the energy distribution of scattered photons were used to simulate the energy spectrum at the maze entrance. The proposed algorithm uses the Monte Carlo code, FLUKA, to calculate a response function for both detectors. To determine the effects of the pile‐up in the spectra, the Poisson distribution was used, employing the average number of photons per pulse (μ) interacting with the detector. The quantity, μ, was obtained from the ratio of the number of events detected to the number of pulses delivered. The energy spectra at various distances from the maze entrance were measured using NaI and plastic scintillation detectors. From these measurements, the values of µ were calculated, and the pile‐up probability was determined. The FLUKA Monte Carlo code was used to calculate the spectrum at the maze entrance and the response matrices of the NaI and plastic scintillation detectors. The algorithm based on the Poisson distribution was applied to calculate the spectrum. Results: The agreement between the calculated and measured spectra was within the first standard deviation of the variance expected in µ. This agreement confirms that photons at the maze entrance have energies between 30 and 240 keV for a maze with three turns, with an average energy of around 85 keV. After pile‐up correction, the range of the pulse height distribution with the plastic scintillation detector, which has a low atomic number, was decreased (0 to 140 keV). In contrast, the range of the pulse height distribution with the NaI scintillation detector was closer to the photon spectrum (0 to 240 keV). Conclusions: The corrected spectrum demonstrates that using a FLUKA Monte Carlo code and an algorithm based on the Poisson distribution are effective methods in removing the distortion due to the pile‐up in LINAC spectra when measuring with NaI and plastic scintillation detectors. The agreement between the corrected and measured spectra indicates that Monte Carlo modeling can accurately determine the spectrum of a LINAC machine at the maze entrance. [ABSTRACT FROM AUTHOR]

A significant challenge to current and future manned and/or unmanned space missions is due to deep space radiation. An improvement in more realistic (more accurate) simulation models in predicting the effects of radiation within the spacecraft is required, especially to better predict dose to astronauts, energy deposition within sensitive electronics, and effectiveness of radiation shielding for long-term space missions. The International Space Station provides an invaluable resource for long-term measurements of the radiation environment in Low Earth Orbit (LEO); however, the only manned missions with dosimetry data available beyond LEO are the Apollo missions. Thus the physiological effects and dosimetry for deep space missions are not well understood in planning extended missions. GEANT4, a Monte Carlo method, represents a powerful physics simulation tool to assess the effects of radiation transport through spacecraft. The newest version of GEANT4 supports multithreading and Message Passing Interface (MPI) allowing for much faster distributive processing of simulations, using a high-performance computing environment. This paper introduces a new simulation model and application using GEANT4 that greatly reduces computational time to hours instead of weeks without any post simulation processing based on high-performance computing. This paper also introduces a new set of GEANT4 computational detectors for calculating dose distribution, besides the historically used International Commission of Radiation Units simulation spheres. The computational detectors include a thermoluminescent detector, tissue equivalent proportional counter, and human phantom, along with additional new scorers to calculate dose equivalence based on the International Commission of Radiation Protection standards. This study presents GEANT4 simulations of the dose deposition for the International Space Station and the Apollo 11 and 14 missions, which replicate well the dose measurements during these missions. The simulations of both Apollo missions show consistent doses from galactic cosmic rays and radiation belts with a small variation in dose distribution across the Apollo capsule. The greatest contributor to radiation dose for both Apollo missions in the simulations came from galactic cosmic rays. Simulations of historical solar particle events during an Apollo missions show a solar particle event would not be fatal and below mission limits. These GEANT4 models also provides the values of the dose deposition and dose equivalent for various organs within a human phantom in the International Space Station and Apollo command module, which are developed for the first time using this GEANT4 based application. • A new application simulated radiation through full-sized spacecraft. • Application implemented new computational detectors including human phantom. • Simulations of dose for ISS and Apollo 11 matched well experimental measurements. • Greatest contributor to radiation dose for Apollo missions was from GCR. • Simulations of Apollo 14 showed higher measured dose most likely due to SPE. [ABSTRACT FROM AUTHOR]

• The selection of the best calculation code is important in radiation protection. • Broder's formula is a good choice for simple geometries with multilayer shields. • The differences between the VISIPLAN and MCNP is mainly caused by Build-up factor. • If the geometry is more complex, use a Monte Carlo code, for example MCNP. Calculations for individual scenarios are carried out for optimization of radiation protection, waste management, decommissioning planning, management of disposal facilities, and others. Comparative study of various calculation codes for different geometries with multilayer shields was performed. Reference geometries are models of the most frequently used waste packages in the decommissioning of nuclear power plants (NPP) in Slovakia. As representatives of deterministic calculation codes were chosen Visiplan, Microshield and Mercurad. Stochastic codes were represented by MCNP code. The impact of different build-up factor (BUF) selection in deterministic codes was evaluated. Experimental measurements with different multilayer shields were performed and compared to the results obtained by MCNP calculations. The percentage deviation between experiment and MCNP calculations was ±20% for every shielding configuration. Analysis of the differences between results was made. Recommendations for choosing of the most suitable calculation code for given application are formulated in this study. [ABSTRACT FROM AUTHOR]

Compact Proton Therapy Centers, CPTC, have a single treatment room, and are technologically more affordable, smaller, advanced and easier to use. From a radiological protection point of view, the leading concern in CPTC are interactions of protons with components of the facility and patients that yield a broad emission of secondary particles, mainly high-energy neutrons, up to 230 MeV, and photons. Optimal design of shielding involves theoretical assumptions in the design phase and, consequently, experimental measurements with extended range neutron detectors must be carried out in the facility during the commissioning period to verify the design, assumptions and building of the enclosures. There are almost 50 CPTC under construction and planning around the world, hence the improvement of methodologies to verify the shielding and to evaluate the dose to workers and general public in CPTC is a trending issue. The aim of this work was to evaluate and compare the response of two commercial extended range REM meters, WENDI-II and LUPIN-II, for their application in shielding verification and radiation area monitoring in CPTC facilities, by estimating the ambient dose equivalent, H*(10), through the Monte Carlo code MCNP6. The results have been compared with previous works. Likewise, the performance evaluation of these devices in continuous energy neutron field have been carried out, using the AmBe/241 neutron source of the Neutronics Hall (NH) of the Neutron Measurements Laboratory of the Energy Engineering Department of Universidad Politecnica de Madrid (LMN-UPM), through Monte Carlo simulation with the MCNP6 code and experimental measurements. The work is framed into the project Contributions to Shielding and Dosimetry of Neutrons in CPTC. • CPTC involve advanced technologies and smaller footprint and represents the next step in the evolution of proton therapy. • Response of WENDI-II and LUPIN-II was assessed for its application in verification of shielding against neutrons in CPTC. • The benchmark in monoenergetic neutron fields was carried out calculating the ambient dose equivalent H*(10) with MCNP6. • The performance in continuous neutron fields have been carried out with experimental measurements using an AmBe/241 source. [ABSTRACT FROM AUTHOR]