Effective Removal Cross-section Research Articles

Theoretical Investigation of the Influence of Different Heavy Metal Oxides Modifiers on ZnO-Bi2O3-B2O3-SiO2’s Photon- and Neutron-Shielding Capabilities Using the Monte Carlo Method

Radiation has become an essential part in medicine and researchers are constituently investigating radiation shielding materials that are suitable for different medical applications. Glass, due to its properties, has been considered an excellent radiation shield for such applications. One of the most common glasses used as a radiation shield is the ZnO-Bi2O3-B2O3-SiO2 anti-radiation glass. Heavy metal oxides have many desirable properties such as high density, transparency to visible light, stability in air and water, high interaction cross section, high infrared transparency, and good absorption of radiation, which make them desirable to be used as modifiers with anti-radiation glass. Research has been focusing on environmentally friendly shielding material which leads to non-lead modifiers such as Na2O, Al2O3, MgO, TiO2, SrO, Sb2O3, and BaO, which have become more desired than PbO. So far, ZnO-Bi2O3-B2O3-SiO2’s photon shielding properties have been studied experimentally with the addition of BaO at certain energies only. In this work, different heavy metal oxides are added as modifiers to ZnO-Bi2O3-B2O3-SiO2 glass in order to investigate theoretically their effects on the shielding properties of the glass at a wide range of photon and neutron energies. Simulation is cost- and time-effective when it comes to investigating different compositions of glass and different modifiers with different weight percentages at any energy range for any type of radiation. Simulation could be considered the first step in order to identify the best mixture with the best weight fractions prior to any experimental investigations of other desired properties based on the needed application. In this work, the photon- and neutron-shielding capabilities of the ZnO-Bi2O3-B2O3-SiO2 anti-radiation glass is investigated with different weight fractions of heavy metal oxides at wide photon and neutron energy ranges. Geant4, which is a Monte Carlo-based powerful toolkit, is used to find the mass attenuation coefficients (µm) of photons, as well as the effective removal cross sections (ΣR) of neutrons, of all the investigated samples in the studied energy range.

Synthesis and Characterization of Zinc-Lead-Phosphate Glasses Doped with Europium for Radiation Shielding

Appropriate glass systems can provide efficient transparent radiation shielding. The current study involved the preparation of the glass system with a composition of xEu2O3-(15-x)ZnO-10CaO-35PbO-40P2O5 (where x = 0, 1, 2, 3, and 4 wt.% Eu2O3). The formation of the glass phase was confirmed using X-ray diffraction (XRD). The study analyzed physical and structural parameters, such as optical conductivity (σopt), refractive index (n), and optical band gap (Eg), with the amount of Eu2O3. The findings indicate that the optical band gap increased as the Eu2O3 content increased. Additionally, a decrease in Urbach energy (EU) was observed, suggesting an improvement in the orderliness of the glass. The study also determined various parameters for gamma-ray shielding, including mass attenuation coefficient (μm), effective atomic number (Zeff), and kerma coefficient (k). For neutron shielding characteristics, the macroscopic effective removal cross-section (ΣR) of fast neutrons was calculated.

Li2B4O7–Bi2O3–ZrO4–CaWO4 glass system for neutron protection in neutron applications

In this study, new glass samples with a chemical composition of 40Li2B4O7+xBi2O3+(55–x)ZrO4+5CaWO4 (x = 10, 20, 30, 40, and 50 wt%) were fabricated using the melt quenching-annealing technique. Important neutron shielding parameters, such as the effective neutron removal cross-section, half-value layer, and neutron transmission factor, of all the new glass samples were calculated using the GEANT4 Monte Carlo simulation code. In addition, neutron dose measurements were performed using an 241Am–Be fast neutron source with an activity of 74 GBq and an average neutron energy of 4.5 MeV and a BF3 gaseous neutron detector. The results were compared with those of conventional shielding materials to determine the neutron absorption capabilities of the new glass materials. All the new glass samples were found to have excellent neutron shielding abilities. In particular, the GB5 sample had a high absorption capacity compared to the other samples. These new glass samples may be used as shielding materials in various radiation applications, such as generation of nuclear energy, radiation medicine, and the transport and storage of radioactive wastes.

Eco-friendly transparent glass prepared from rice straw ash for neutron and charged particle radiation shielding

The current study aimed to fabricate and investigate new glass system incorporating silica from rice straw ash (RSA). The effect of adding different concentrations of bismuth oxide (Bi2O3) on the glass's structure was examined using FTIR. The A.C. conductivity of the prepared glasses over the frequency range (120 Hz − 100 kHz) and the temperature range (300–620 oK) have been measured. Based on its electron density, a substance's electrical conductivity can alter as photon energy increases. For the current investigated glasses, the electron density (Neff) and electrical conductivity relation over the energy range (0.015–10 MeV) were calculated. Glass samples' ability to attenuate radiation such as neutrons, electrons, protons, and alpha particles has been investigated. For neutrons, the effective removal cross-section (∑R, cm−1) and the total macroscopic cross-section (∑T, cm−1) are calculated. For charged particles like protons (H+) and alpha (He2+), SRIM software has been used to determine the glass Mass Stopping Power (MSP) and the Projected Range (PR). For electrons (β−), MSP and the Continuous Stopping Down Approximation (CSPA) were calculated using ESTAR (NIST). On the other hand, Phy X/Zextra software was used to determine glasses effective atomic numbers (Zeff) over a broad energy range. As a result, BSiBi20 glass with the highest Bi2O3 concentrations (20 mol. %) shows superior shielding features for fast neutrons (∑R = 0.128 cm−1) compared to the competitors like ordinary concrete (O.C.). For charged particles, BSiBi20 glass achieves the lowest MSP, CSPA, and the highest Zeff values over the entire energy range. The KERMA (Kinetic Energy Released per unit Mass of the Absorbing material) calculations indicated a significant influence on the elemental composition of the glass system in the intermediate photon energy regime.

Radiation shielding performance of metal oxides/EPDM rubber composites using Geant4 simulation and computational study

This paper aimed to evaluate the shielding performance of ethylene propylene diene monomer (EPDM) rubber composites filled with 200 phr of different metal oxides (either Al2O3, CuO, CdO, Gd2O3, or Bi2O3) as protective materials against gamma and neutron radiations. For this purpose, different shielding parameters, including the linear attenuation coefficient (μ), mass attenuation coefficient (μ/ρ), mean free path (MFP), half value layer (HVL), and tenth value layer (TVL), were calculated in the energy range between 0.015 and 15 MeV by using the Geant4 Monte Carlo simulation toolkit. The simulated μ/ρ values were validated by the XCOM software to examine the precision of the simulated results. The maximum relative deviation between the Geant4 simulation and XCOM was not greater than 1.41%, confirming the accuracy of the simulated results. Based on μ/ρ values, other significant shielding parameters such as effective atomic number (Zeff), effective electron density (Neff), equivalent atomic number (Zeq), and exposure buildup factor (EBF) were also computed to explore the potential usage of the proposed metal oxide/EPDM rubber composites as radiation protective materials. The study demonstrates that the gamma-radiation shielding performance of the proposed metal oxide/EPDM rubber composites are increasing in the order of EPDM < Al2O3/EPDM < CuO/EPDM < CdO/EPDM < Gd2O3/EPDM < Bi2O3/EPDM. Furthermore, three sudden increases in the shielding capability in some composites occur at 0.0267 MeV for CdO/EPDM, 0.0502 MeV for Gd2O3/EPDM, and 0.0905 MeV for Bi2O3/EPDM composites. This increase in the shielding performance is due to the K absorption edges of Cd, Gd, and Bi, respectively. Regarding the neutron shielding performance, the macroscopic effective removal cross-section for fast neutrons (ƩR) was evaluated for the investigated composites using MRCsC software. The highest ƩR is obtained for Al2O3/EPDM, while the lowest ƩR is obtained for EPDM rubber with no metal oxide content. According to the obtained results, the investigated metal oxide/EPDM rubber composites can be employed as comfortable clothing and gloves designed for workers in radiation facilities.

Investigation of the radioprotective properties of Zingerone, Ibuprofen, and Sulfasalazine against neutron radiation damage

Nuclear radiation (neutron, gamma X rays, etc.) is widely used in power generation, industry, medicine, research, and agriculture. Exposure to ionizing or indirect ionizing radiation can cause mutagenetic cell death depending on the dose and duration. It is important to reduce the carcinogenesis of cancer patients undergoing radiotherapy treatment and people working in radiation fields. Neutrons have wide applications in neutron activation analysis (NAA), neutron capture therapy (NCT), neutron radiography (NR), neutron capture, neutron conversion doping (NTD) sample irradiation, and neutron diffraction (ND). In these applications, protective substances are needed to stop the neutrons from entering the body and to reduce their effects. In this work, the use of zingerone, ibuprofen, and sulfasalazine compounds as active ingredients in a neutron protective drug was investigated. Some important neutron attenuation parameters such as an effective removal cross section, mean free path, half-value layer, and transmission number were theoretically calculated with the Monte Carlo simulation GEANT4 code for fast and epithermal neutrons.The 241Am-Be fast neutron source and a BF3 gaseous detector were used to determine the neutron absorption ability of all of the samples. All of the results were compared with paraffin, which has a good shielding capacity. According to the results, it was found that every examined sample had a fast and epithermal neutron absorption capacity. It reported that the dose amount from the source was absorbed by the samples at rates of 33.37% (DAI1), 34.85% (DAI2), and 34.21% (DAI3). It was determined that the DAI2 sample in particular had a high absorption ability ratio compared to other samples. Because this sample have high hydrogen content than others samples. So, it was shown that these compounds can be used as active substances to make a radioprotective drug.

Assessment of Five Concrete Types as Candidate Shielding Materials for a Compact Radiation Source Based on the IECF

A radiation source based on the inertial electrostatic confinement fusion (IECF) system is being developed for multidisciplinary research applications. The radiation outputs from the IECF system are 2.45 MeV fast neutrons and the associated co-generated X-rays with an energy less than 3 MeV. A radiation shielding study has been performed on five types of concrete to define the most efficient material for the shielding design of the system. The proposed materials were ilmenite-magnetite concrete (IMC), ordinary concrete-1 (OC-1), barite-containing concrete (BC), ordinary concrete-2 (OC-2), and serpentine-containing concrete (SC). A numerical model was applied to determine the effective removal cross-section coefficients (∑Rt) for the fast neutrons and the total mass attenuation coefficients (µm), the half-value layer (HVL), the mean free path (MFP), the effective atomic number (Zeff), and effective electron density (Neff) for photons inside the materials. The model considered the radiation source energy and the material properties of the concrete types. The results revealed that the serpentine-containing concrete exhibited the highest ∑Rt with 12 cm of concrete thickness needed to attenuate an incident neutron flux to 1/100 of its initial value. In addition, the BC shows the highest µm with a 38 cm concrete thickness needed to attenuate the 3 MeV energy X-ray flux to 1/100 of its initial value. This study suggests that a 40 cm thickness of SC or BC adequately shields the radiation generated from an IECF system with a maximum particle production rate of up to 1 × 107 n/s.

Investigation of the neutron radiation protective properties of chlorophyll and carotenoid

In this study, we have investigated neutron radiation absorption properties of chlorophyll and carotenoid. Our aim is to show that chlorophyll and carotenoid, which are naturally found in every green plant, can be used to make a cream, medicine, paint, or shield material against neutrons. Neutron radiation is generally used in many applications such as diffraction and scattering experiments, material crystallography, diagnosis and therapy in medicine, cell DNA studies in biology, condensed matter physics, materials research and development, solid-state chemistry, mineralogy, geology, and the study and improvement of the structural properties of materials. In order to reduce or completely eliminate the damages that can be caused by direct or scattered neutrons by the people working in these studies or other people in the environment, materials with high shielding are needed to absorb neutrons. In order to understand that chlorophyll and carotenoid can be used in the production of neutron shielding materials, the neutron radiation absorption parameters of these materials were investigated. The parameters, which are widely used in neutron shielding studies such as an effective removal cross section, half value layer, mean free path, and neutron transmission factor (NTF) were estimated using the Monte Carlo simulation GEANT4 code for both fast and epithermal neutrons. In addition, the dose rates of all samples were determined by using an 241Am–Be fast neutron source. All theoretical and experimental results obtained to understand the absorption capacities of chlorophyll and carotenoid were compared with paraffin, which is widely used in neutron shielding applications.It is to be noted that determined that chlorophyll showed an effective shielding against fast neutrons and carotenoid showed an effective shielding against epithermal neutrons. Investigation revealed that these materials are alternate shielding materials against neutron radiation for the production of medicine, composite material, and paint to be developed for neutron radiation.

Effect of Ce-dopant on photon interaction parameters of zirconolite-based borosilicate glass ceramics

In the present investigation, 70 wt% of zirconolite (CaZr1-xCexTi2O7, x = 0, 0.2, 0.4, 0.6, 0.8 and 1) and 30 wt% of sodium borosilicate glass (NaB0.5Al0.5Si2O6) (sample C0 to C10) have been synthesized. The photon interaction parameters viz. mass attenuation coefficient, effective atomic number, effective electron density, mean free path, and half value layer have been investigated for glass-ceramic samples at incident photon energies of 59.58 keV, 81.2 keV, 356 keV, 662 keV, 1173 keV and 1333 keV. Furthermore, effective neutron removal cross-section values, mass stopping power (MSP) and range of alpha particles and proton, were also calculated using SRIM codes. Moreover, gamma-ray energy absorption buildup factor (EABF) and exposure buildup factor (EBF) were also estimated. The experimental results are in agreement with the theoretically calculated values using WinXCOM program. Among the given prepared samples, C10 shows the best shielding properties, and C0 shows the best capability for neutron absorption. The MSP and projected range of α-particles and protons in all the prepared samples show very small variation with the change of Ce content in zirconolite in the energy range upto 15 MeV. The measured values are extensively desirable in fission and fusion reactors, medical applications, shielding researchers, and radiation physics.

ENDF/B-VIII.0-based fast neutron removal cross sections database in Z = 1 to 92 generated via multi-layered spherical geometry

A fast neutron mass removal cross sections (ΣR/ρ) database covering the elemental range of Z = 1 to 92 is generated for source neutron energy distributions based on 241Am–Be, 252Cf, and 235U. Monte Carlo simulations using ENDF/B-VIII.0 nuclear data were performed with the fast neutron removal theory's ideal spherical representation. The ENDF/B-VIII.0-based ΣR/ρ′s derived were compared with values from the Oak Ridge National Laboratory's Lid Tank Shielding Facility (LTSF) experiments, values from software Phy-X/PSD and MRCsC, values obtained by an empirical model in Zoller, and values derived using the older ENDF/B-VII.0 nuclear data library. Subsequently, the generated ΣR/ρ′s were evaluated for convergence under varying moderator thickness. Results indicate that the ΣR/ρ quantity for fission and (alpha, n) spectra is universal, with slight deviations for low Z elements. The ΣR/ρ values generated from ENDF/B-VIII.0 agreed with LTSF experiments with an average relative deviation of 8.58%. Moreover, using ENDF/B-VIII.0-based ΣR/ρ′s, the derived fast neutron effective removal cross sections (ΣR's) for several light compounds and dry concrete have better agreement on average with LTSF experiments compared with values derived from the available ΣR/ρ′s in the literature. Significant differences were found for ΣR/ρ′s that were generated with ENDF/B-VIII.0 as compared with the superseded ENDF/B-VII.0. Furthermore, ENDF/B-generated ΣR/ρ′s through spherical model showed the expected convergence towards singular values at moderator thicknesses practical for fast neutron removal theory applications. Owing to the relatively high unavoidable uncertainties of the LTSF experimental ΣR/ρ′s, the ENDF/B-VIII.0-based ΣR/ρ′s present a new and potentially favorable database in ΣR general calculations for homogeneous multi-element materials for point-kernel shielding applications.

Investigation of the gamma and neutron radiation properties of strontium bismuth borate glasses doped with neodymium ions using the Geant4 simulation toolkit and EpiXS software

In this study, the shielding capabilities of glasses made of B2O3 Bi2O3 SrO Nd2O3 have been investigated for use in radiation protection applications. Therefore, to simulate the mass attenuation coefficient for the glass samples A, B, C, D, E, and F, we utilized the Monte Carlo simulation Geant4 code. The findings of the simulated mass attenuation (µm) were compared with those that the EpiXS software had theoretically predicted for photon energies between 30 and 2000 KeV. The simulated µm findings and the EpiXS received data were found to be in good agreement. Based on the simulated µm, it was possible to determine the linear attenuation coefficients (µ), effective atomic number (Zeff), effective electron density (Neff), half-value layer Δ0.5, and mean free path (MFP). With an increase in photon energy and a rise in the (mol percent) proportion of Nd2O3, the µm value fell. Due to its greater values of the linear attenuation coefficient, effective atomic number, and lower values of Δ0.5 and MFP, it is discovered that the F glass with 5.7 (mol. percent) concentration of Nd2O3 has superior gamma-ray shielding effectiveness. Additionally, the effectiveness of the chosen glasses in neutron shielding was assessed in this paper. The Monte Carlo simulation Geant4 code was used to determine the effective removal cross-section (cm−1). As a result of having the highest effective removal cross-section and the lowest HVL at energy 0.5 MeV, it is determined that the F Glass with a 47.2 (mol. percent) concentration of B2O3 and 5.7 (mol. percent) concentration of Nd2O3 has higher neutron shielding qualities than the other glass samples that were chosen.

Ionized and non-ionized radiation effects on coronary stent implantation

The purpose of this study is to examine the clinical radiation effects on patients with various stents. Several previous researches focused on the mechanical and physical features of the stents, but there have been few studies that focus on the interaction with radiological and clinical radiation. For stent material analysis, the ANSYS package program was employed. These materials and models are often built in three dimensions for three different types of stents made of three various materials. Estimates of blood pressure and thermal radiation were explored, as were the effects of non-ionizing radiation. Furthermore, the radiation attenuation characteristics of stent samples are investigated. The mass attenuation coefficient values are computed using MATLAB code over a large energy range of 0.015–15 MeV, and the findings are validated using theoretical WinXCom results. To determine the gamma-ray attenuation performances of the studied stent samples, variables such as the mass attenuation coefficient (MAC), half-value layer (HVL), tenth-value layer (TVL), mean free path (MFP), effective atomic number (Zeff), exposure build-up factor (EBF), and energy absorption build-up factor (EABF) are computed. Effective removal cross-sections (ΣR) of stent samples were acquired to determine the capacity of stent samples to stop fast neutrons. Finally, the ability of stent samples to stop charged alpha and proton particles was evaluated utilizing mass stopping power and projected range parameters. The discovery demonstrates that S3 has the best attenuation as well as the best proton, alpha, and gamma radiation attenuation capability.

Comparative shielding features for X/Gamma-rays, fast and thermal neutrons of some gadolinium silicoborate glasses

X/Gamma-rays, fast and thermal neutrons shielding properties for eight gadolinium based glasses (labeled as S1:S8) have been investigated at energies between (0.015–15 MeV) using MCNPX code and WinXcom software. The correlation coefficient (R2) was used to assess the degree to which MCNPX results are related to that of WinXcom. For all glass samples, R2 was found close to 1, this infers a perfect degree of correlation between the MCNPX and WinXcom results. Multiple parameters including, mass attenuation coefficient (μ/ρ), half value layer (HVL), mean free path (MFP), effective atomic number (Zeff), electron density (Nel), radiation protection efficiency (RPE), exposure and energy absorption buildup factors (EBF & EABF), specific absorbed fraction of energy (SAFE), kerma relative to air (Ka), macroscopic effective removal cross section (ΣR), and total neutron cross-section of thermal neutron (σt) were determined and compared with that of other frequently used along with recently studied shielding materials. The results show that, as the Gd2O3 content increases, the μ/ρ, Zeff, Nel, RPE, ΣR and σt values increase while HVL, MFP, EBF, EABF, and SAFE values all decrease as follow: the values of MAC ranges from 0.0173 to 1.906 and 0.0281–31.831 cm2 g−1; Zeff varied from 6.027 to 8.836 and 10.497–45.659; HVL varied from 0.152 to 16.750 and 0.005–5.533 cm; Nel varied (from 2.8 to 4.25) x1023 and (2.88–13.96) x1023 (electrons/g) for S1 and S8 respectively. In addition, among all studied glasses, S8 glass sample (with G2dO3 content of 35% mol) provided a fully protection for X-rays (RPE = 100%). Moreover, when compared to other common shielding materials and recently studied glasses (in terms of MFP), the S8 sample was found to have the highest attenuation performance relative to many of these materials. In terms of fast and thermal neutrons, the addition of Gd2O3 had a great effect on increasing the attenuation capability for fast and thermal neutrons as follow: the ΣR and σtotal values were increased from 0.120 to 0.145 cm−1 and from 123.99 to 15,258.73 b for S1 and S8 respectively. In addition, the S8 sample was found to have the highest neutron attenuation performance when compared to other common shielding materials and recently studied glasses. Therefore, S8 sample found to have superior characteristics for radiation and neutron shielding applications. The findings of this study might be highly useful in the future for medicinal and industrial applications.

Experimental and theoretical investigation on physical, structure and protection features of TeO2–B2O3 glass doped with PbO in terms of gamma, neutron, proton and alpha particles

Glass series of composition 70TeO2- (30-x)B2O3- xPbO (where x = 2, 6, 10, 14 and 20 mol%) is prepared using melt-quenching technique. The glass structure was investigated by X-ray diffraction (XRD). Manufactured glass density, molar volume and Fourier transform infrared spectroscopy (FTIR) were measured. Shielding parameters against gamma-rays including, linear attenuation coefficient (μ), mass attenuation coefficients (μm), half-value layer (HVL) and mean free path (λ) were measured experimentally at energies of 0.664, 1.177 and 1.334 MeV using Cs137 and Co60 point sources and NaI(TL) scintillation detector. The measured parameters were compared with those calculated using XCOM program with a good agreement between them. To fully comprehend the shielding efficiency of the prepared glasses, several shielding parameters including, μm, HVL, effective atomic number (Zeff), electron density (Nel), equivalent atomic number (Zeq), exposure and energy absorption buildup factors (EBF & EABF), radiation protection efficiency (RPE) and kerma relative to air (Ka), were evaluated in the photon energy range 0.015–15 MeV and compared with some materials that commonly used as gamma-ray shielding. The findings demonstrate the viability of employing the current glass composition as an effective agent for gamma-ray shielding and fully protection (100%) for x-ray. The macroscopic effective removal cross section (ΣR) of the prepared glasses were calculated and compared with those of frequently used neutron shielding materials. Finally, the mass stopping power (MSP) and projected range (PR) values of proton (H1) and alpha particles (He+2) have been evaluated in the energy range 0.01–15 MeV. The results show that, increasing PbO content improves shielding characteristics. Among the examined samples, the S5 glass sample was found to be superior in terms of photon, neutron, proton and alpha particles protection. These results indicate that, the prepared glass samples are highly appropriate nuclear shielding material that can be used in many applications.

Determination of biological radioprotective characteristics of some natural organic compounds for radiation shielding applications

Abstract Ionizing radiation is used in many fields in energy, medicine, and industrial applications. Those who are in these areas or cancer patients receiving radiotherapy are at risk for acute or long-term exposure to radiation damage due to these ionizing radiations. Non-toxic new agents are needed to protect intact tissue and cells. In this study, we aimed to determine the gamma and neutron radiation attenuation characteristics of seven different natural compounds (quercetin, menadione, naphthol, caffeine, quinine sulphate, cholesterol and riboflavin) to help users in radiation applications. Gamma radiation attenuation parameters such as the mean free path, mass attenuation coefficient, effective atom number, linear attenuation coefficient, and half-value layer were calculated theoretically with WinXCom software for the energy range 0.015–15 MeV. Fast neutron attenuation criteria, such as mean free path, half-value layer, effective removal cross-sections and transmission neutron number, were theoretically determined with Monte Carlo simulation codes (Geant4). Neutron absorption measurement experiments were also applied in addition to the theoretical results. The neutron radiation absorption capacities were determined for samples with an 241Am-Be 4.5 MeV energy neutron source and portatif-type Canberra brand BF3 gas neutron detector. Neutron attenuation parameters were compared with paraffin to determine the absorption capability of the samples. It was found that the dose 1.1094 (μSv/h) from the source was absorbed by the samples to the following extent: 31.76% (Quercetin), 21.85% (Menadione), 28.85% (Naphthol), 22.94% (Caffeine), 12.51% (Quinine sulphate), 40.44% (Cholesterol) and 20.94% (Riboflavin). From the results, it can be clearly seen that all these drug samples had a good neutron radiation attenuation capacity. This revealed that the examined samples had radiation absorption abilities. It was found that the cholesterol sample had an especially excellent absorption power for both neutron and gamma radiation. The samples investigated in this study could be used to develop radiation-protective drugs.

Theoretical Investigation of Gamma- and Neutron-Shielding Properties of Polysulfone (PSU) Polymer Material Using Geant4.

Polymers are widely used materials that have many medical and industrial applications. Some polymers have even been introduced as radiation-shielding materials; therefore, many studies are focusing on new polymers and their interactions with photons and neutrons. Research has focused on theoretical estimation of the shielding effectiveness of different materials. It is well known that theoretical studies on the shielding properties of different materials through modeling and simulation have many benefits, as they help scientists to choose the right shielding material for a specific application, and they are also much more cost-effective and take much less time compared to experimental studies. In this study, polysulfone (PSU) was investigated. PSU is a high-temperature, amber-colored, semi-transparent plastic material with good mechanical properties. It is resistant to degradation from hot water and steam and is often used in medical and food preparation applications, where repeated sterilization is required. The interactions of photons and neutrons with PSU were investigated using a Monte Carlo-based simulation toolkit, Geant4, within a wide range of energies of both photons and neutrons. The mass attenuation coefficients (µm), the half-value layers (HVL), the effective atomic numbers (Zeff), and the effective electron densities (Neff) of gammas were investigated. In addition, the effective removal cross-sections (ΣR) and the mean free paths (λ) of neutrons were also studied. The results were then compared to other commonly used polymer materials.

Gamma and Neutron Shielding Parameters of Polyester-based composites reinforced with boron and tin nanopowders

The usage of composites as the shielding materials are highly recommended since they could be used in order to attenuate the undesired radiation with unique properties and advantages in the areas where the radiation is prevalent. In this context, not only are their radiation shielding properties important but also their flexibility, durability and low cost. Due to the mentioned superior characteristics, the polyester based composites are among the most preferred materials. With the aim of creating unique and novel radiation shielding materials, this study investigates gamma and neutron shielding capabilities of the polyester composites reinforced with Boron and Tin nanopowders at different proportions (0–50%, 10–40%, 20–30%, 30-20% and 40-10%, 50-0%). The gamma shielding abilities of the prepared polyester composite materials were evaluated using an HPGe detector system, WinXCOM computer program and different simulation tools (FLUKA and GEANT4) at the energies varying from 59.5 to 1332.5 keV. The experimental, theoretical and simulation results showed remarkable agreement between each other, and the addition of Sn enhances the gamma attenuation performance of the chosen polyester composite materials. In addition to gamma analysis results, neutron shielding properties of the proposed composites are further determined. On this purpose, the transmitted neutron numbers through the samples (as functions of neutron energy and the sample thickness) and effective neutron removal cross sections were evaluated. The neutron shielding performance of the samples showed that the prepared composites could be alternative materials to the existing neutron shields in the literature.

Structural, thermal and nuclear shielding properties of lithium lead phosphate glasses: Influence of lithium oxide additive

In this paper, the structural, thermal and nuclear radiation shielding properties of a new Lead–Lithium-Phosphate glass system with different amount lithium oxide were studied. The glass samples were prepared by the conventional melt-quenching method and they having compositions (75-x) P2O5– 25PbO- xLi2O with x = 0, 5, 10, 15, 20 mol%. The influence of Li2O content on the structural, thermal, radiation shielding properties, and other related properties of the synthesized glasses was studied. The structural properties were investigated using XRD and Raman spectroscopy. The glass transition temperature Tg and crystallization onset temperature Tx are estimated from the DSC profile.The GEANT4 Monte Carlo simulation code was used to estimate thermal neutron attenuation coefficient (μnt) in a wide neutron energy range of 0.0253–100 eV and ParShield code was used to calculate effective fast neutrons removal cross-section (∑R). The gamma-ray attenuation parameters such as mass attenuation coefficients (μm), effective atomic number (Zeff), effective electron density (Neff) and half-value layer (HVL) have been computed by using ParShield code in the energy range of 0.103–1.33 MeV. However, the total mass stopping power (MSP) and projected range (PR) values have been estimated for alpha particles (4He2+) and Triton particle (3H) by using SRIM in the energy range of 0.1–10 MeV.

Investigation of radiation protective features of azadispiro derivatives and their genotoxic potential with Ames/Salmonella test system

Purpose Five different types of synthesized azadispiro derivatives have been analyzed for radiation absorption capacity and determined their potential to be exploited as substances for a drug to be developed against radiation has been investigated. Material and methods Fast neutron attenuation parameters like the effective mean free path, half-value layer (HVL), removal cross-sections, and neutron transmission number were found with the Monte Carlo simulation Geometry And Tracking (GEANT4) code. Gamma radiation absorption parameters, such as effective atom number (Z eff), mean free path (MFP), mass attenuation coefficient (MAC), and half-value layer (HVL) were theoretically determined with WinXCom software. Besides, the exposure build-up factor (EBF) was calculated by using GP fitting parameters. Neutron absorption dose rate was experimentally calculated with 241Am-Be fast neutron source which has 4.5 MeV of energy, 74 GBq activity, and portative BF3 neutron detector. Ames/Salmonella test systems were used for the genotoxic potentials of the azadispiro derivatives Results and conclusions Experimental and theoretical results were checked with paraffin and High-Density Polyethylene. The results showed that Azadispiro derivatives have neutron radiation absorption capability close to paraffin and High-Density Polyethylene. The gamma radiation absorption properties for azadispiro derivatives have been investigated, and it has been observed that these materials can absorb gamma radiation. Ames/Salmonella assay was used to examine whether the derivatives had a genotoxic effect probability or not. The results showed that these derivatives were genotoxic and safe at test doses (up to 5 mM). Consequently, it has been understood that these azadispiro derivatives can be used as active and genotoxic safety ingredients in the production of a protective drug against both neutrons and gamma rays.

Gamma ray and fast neutron shielding of ZrSiO4-Al2O3 ceramic refractor

The aim of the present research is to investigate the role of three composites samples based on ZrSiO4/Al2O3 ratios (ZA91, ZA73, and ZA55) on the attenuation of gamma rays and fast neutrons. The sintering of the samples at 1450 °C/3h show the formation of composite of zircon and corundum and their microstructure varying from irregular and directed rods and scattered particles in ZA91 to clear quadrant euhedral crystals in ZA55. The heavy metal (ZA91) designated higher attenuation of gamma rays compared to ZA73 and ZA55 using photon energy of 0.662, 1.173, and 1.332 MeV, which may be attributed to the higher density of ZA91 (4.1910 g/cm3). In contrast, the attenuation property of light metal (ZA55) for fast neutrons is higher related to other composites. This may be correlated to the high interaction scattering cross-section of alumina content compared to that of zircon. The microstructure and mineralogical features were analyzed by scanning electron microscopy and x-ray diffraction (XRD), respectively. In addition to the gamma transmission parameters of linear attenuation coefficients (µ) and half value layer of the composite for the radio cesium (Cs137) and cobalt (Co60) as gamma source are measured. Likewise, the effective removal cross-section of fast neutrons (ΣR) spectra by composite for californium radioisotope (Cf252) as neutrons source was deliberated. Hence, it could be decided that the optimum ratios of composites based on zircon (ZrSiO4) and corundum (Al2O3) will be governed by the type of radiation emitted by the nuclear facility.