Neutron Attenuation Research Articles

FTIR, Raman spectroscopic, microstructure and neutron-particle interaction properties of Mo3+ doped cadmium zinc lithium-borate glasses

The application of different glass materials in the area of radiation shielding and charge particle attenuation has necessitated a comprehensive study of glass materials doped with special elements to examine the impact of the special element on the microstructure and shielding capabilities of resultant glass materials. In this study, the optical absorption spectra of Mo3+ doped cadmium zinc lithium borate glasses (BCZLMx) was investigated using X-ray diffraction (XRD), Fourier transform infrared (FTIR) spectroscopy and Raman spectroscopic methods to ascertain the suitability of BCZLMx glass microstructure to neutron and charge particle attenuation. Results show that the absorption cross-section is the dominant contributor to the total cross section for all samples, ranging from 110.9 cm−1 to 111.7 cm−1. BCZLM1 glass sample appears to be slightly more effective at removing fast neutrons from the beam while BCZLM5 glasses are effective at stopping electrons, protons, alpha particles, and carbon ions when compared to the other BCZLMx glasses samples. This implies that doping cadmium zinc lithium borate glasses with Mo3+ reduces fast and thermal neutrons shielding potential and increased charge particles interaction of the BCZLMx glass system.

Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300μSv/h and 175-200μSv/h, respectively, when the reactor is operating at 10kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178keV. Apart from energy reduction, 10kW normalised neutron flux of south beam pit is also attenuated by the shield by 25000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

Structural, Morphological, and γ-ray Attenuation properties of m-type hexaferrite BaFe12O19 doped with V2O5, Ce2O3 and Bi2O3 for Radiation Shielding applications

This paper studies, structurally analyzes, evaluates the morphology, and assesses the gamma-ray and neutron attenuation properties of a composite material composed of M-type hexaferrite BaFe12O19 (BF) doped with three additional compounds: Bi2O3, V2O5, and Ce2O3. X-ray diffraction revealed that the structure of BF can be indexed to a hexagonal structure, BF/Bi2O3 to a tetragonal structure, BF/V2O5 to a cubic structure, and BF/Ce2O3 to a trigonal structure. These results were refined using PROFEX software. The Crystallite size of the nano-powders was determined by X-ray line broadening using the Williamson-Hall method. Morphology evaluations illustrated that HR-TEM images for BF, BF/Bi2O3, BF/V2O5, and BF/Ce2O3 reveal high crystallinity with a polycrystalline nature. Various attenuation measurements, using the FLUKA Monte Carlo code, aimed to determine the efficacy of these dopants into m-type hexaferrite BF as potential materials for radiation shielding applications. Doping with Bi2O3 into hexaferrite BF markedly enhances its attenuation properties, demonstrating superior Linear Attenuation Coefficient (LAC) values across a broad energy spectrum from 0.16 to 10 MeV. The Mass Attenuation Coefficient (MAC) analysis revealed a pivotal energy threshold at 0.3 MeV. Below this energy, BF presented the highest MAC values. Conversely, for energies equal to or greater than 0.3 MeV, BF exhibited the highest MAC values, affirming its superior gamma photon shielding efficacy. The study of the Half Value Layer (Δ0.5) and Mean Free Path (MFP) provided additional evidence of the superior shielding performance of BF/Bi2O3. The Δ0.5 values ranged between 0.128 to 1.240 cm for BF, 0.162 to 1.430 cm for BF/V2O5, 0.163 to 1.390 cm for BF/Ce2O3, and 0.162 to 1.210 cm for BF/Bi2O3, indicating that BF/Bi2O3 requires a thinner layer to halve the gamma-ray intensity, particularly at higher energies. The MFP values further support the improved shielding ability of BF/Bi2O3, with measurements showing that gamma photons travel the shortest distance before interacting with the material, thereby reducing the potential for radiation exposure.

Effect of fibers and boron carbide on the radiation shielding properties of limestone and magnetite aggregate concrete

Research on new composite materials in the field of radiation shielding materials requires adequate means and meticulousness in their manufacture and characterization. In this study, concretes that use elements that theoretically improve their shielding properties are developed, such as fibers and boron-rich additions. Both magnetite and limestone concrete have been manufactured, incorporating steel fibers to improve the linear attenuation coefficient and polyvinyl alcohol to increase neutron scattering cross section. The incorporation of boron carbide is intended to increase the neutron absorption cross section. 5 mixes have been manufactured that have been tested with several gamma radiation sources (570–1280 keV) and with an Am–Be neutron source (111 GBq). The experimental results have been used to validate simulations carried out with MAVRIC® and have led to the modelling of a simplified spent nuclear fuel cask. Boron carbide produces a decrease of more than 50% on the neutron dose rate in contact with the cask. Incorporating PVA fibers supposes gain in neutron attenuation capacity of up to 13%, while fibers have a negligible effect on the photon dose rate.

Neutron Shielding Properties of Cellulose Acetate CdO-ZnO Polymer Composites

In this work, the neutron shielding ability of Cellulose Acetate-CdO-ZnO Polymer Composites of different concentrations of CdO and ZnO were investigated. Cellulose acetate is a biodegradable good matrix and the used metal oxides are good for absorbing radiation. The neutron attenuation coefficient was calculated by Phy-X computer code for all the samples.

Density Measurements of Molten LiF-BeF2 and LiF-BeF2-LaF3 Salt Mixtures by Neutron Radiography.

The densities of eutectic (LiF)2-BeF2 and mixtures of this salt (FLiBe) with LaF3 were measured by dilatometry and by neutron attenuation from 673 K to 1,073 K. Because LaF3 has a limited solubility in FLiBe, it was necessary to determine the amount of LaF3 in solution before the density could be determined. The FLiBe density determination was favorably benchmarked against the literature data. A simple comparison was not available for the LaF3-FLiBe mixtures, so extrapolation of published data was necessary based on analysis using the Molten Salt Thermal Properties Database-Thermochemistry, or MSTDB-TC, developed by the US Department of Energy. Solubilities for LaF3 in FLiBe ranged from 1 to 4 mol % over 673 to 1,073 K. The salt system was heated and cooled over 24 h to evaluate potential changes in composition and hysteresis during the measurement. Changes in the meniscus were observed, and these were included in the correction for density determinations. Salt surface tension may have led to supersaturation of LaF3 in the salt because the solubility curve was nonlinear with respect to the inverse temperature, as would be expected for an ideal system. Surface tension measurements are currently underway to test this hypothesis.

Strontium Oxide-Reinforced Borotellurite Glasses: Synthesis, Structure, and Optical Characteristics and γ-Ray and Neutron Attenuation Capability

Strontium Oxide-Reinforced Borotellurite Glasses: Synthesis, Structure, and Optical Characteristics and γ-Ray and Neutron Attenuation Capability

Recycled cathode ray tube waste glasses for radiation shielding applications: Role of Na2CO3

The aim of the study is to evaluate the ability of cathode-ray tube funnel waste glasses (CRT-F) to shield photons (gamma and X-ray) and other forms of radiation with rest masses (i.e., neutrons and charged particles) after their Pb content was reclaimed. The lead (Pb) content of cathode-ray tube funnel waste glasses (CRT-F) was reduced using Na2CO3 as the reducing agent CRT-F(Na2CO3). CRT-F and CRT-F(Na2CO3) were produced using the melt and quench processes using the powder of pristine funnel glass from discarded cathode ray tube glass and the reduced glass powder, respectively, as the starting materials. The X-ray fluorescence technique was adopted for determining the chemical composition of CRT-F and CRT-F(Na2CO3) glasses. The parameters relating to the gamma photon, fissile neutron, moderated neutron, slow neutron, and charged radiation (β, H+, He2+, and C6+) attenuation abilities of pristine CRT-F and CRT-F(Na2CO3) were evaluated. The FLUKA simulation code and XCOM were used to estimate the mass attenuation coefficient of the glasses. The stopping power (Sp) and range (R) of β, H+, and He2+ were evaluated by using the NIST data-based calculators (ESTAR for β, PSTAR for H+, and ASTAR for He2+) while Sp and R data for C6+ was determined using SRIM. The cross section for fast, thermal (25 meV), and slow neutrons (0.1–10 meV) was also estimated. The Pb reclamation from the CRT-F drastically increased the half value layer of photons by at least a factor of 4 at 0.1 MeV and by more than 90 % at 10 MeV. The CRT-F interacts more with fissile and thermal neutrons compared to CRT-F(Na2CO3). Although, CRT-F showed a better ability to attenuate all the radiation types considered, both glasses are better absorbers of radiation (especially photons) than some common shields. The reduction of Pb weight in CRT-F resulted in a drastic reduction in the ability of the glass to attenuate radiation; the resulting glass (CRT-F(Na2CO3)) is however, an environmentally safer glass shield.

Valorization of steel slag into sustainable high-performance radiation shielding concrete

The disposal of steel slag (SS) has become a serious ecological problem due to the waste of massive land resources. This study reveals the huge potential of recycling SS as heavy aggregates to develop radiation shielding concrete (RSC). The physical characteristics, radiation attenuation capabilities, microstructure, and environmental and ecological impacts of SS-based RSC are comprehensively evaluated and compared with those of normal concrete and the commonly used heavy concrete. The test results indicate that SS-based RSC is prospective to substitute the commonly used heavy concrete, emerging as a sustainable and efficient material for radiation shielding in nuclear energy engineering. Specifically, due to the reaction of cementitious active substances in the surface of SS, the developed SS-based RSC exhibits a fantastic interfacial transition zone (ITZ), and a compact pore structure, thereby showing excellent mechanical properties. The maximum compressive strength of developed SS-based RSC reaches 47.4 MPa. In terms of gamma ray attenuation property, due to the introduction of heavy nuclei and the increase in density, the gamma ray attenuation capacity of developed SS-based RSC is 17.2 % higher than that of normal concrete. The neutron attenuation property is also slightly enhanced due to the refinement effect of SS on pore structure of RSC. Life cycle assessment and unit cost evaluation results demonstrate that compared with barite based-RSC, SS-based RSC can save 97.6 % in cost, 52.0 % in energy consumption, and 70.4 % in carbon emission. The developed SS-based RSC exhibits excellent mechanical properties and radiation attenuation capacities, contributing to the further application of radiation shielding materials. Besides, this approach achieves the high-value recycling of SS, providing new insights for the utilization of high-density solid wastes.

Polyurethane reinforced with micro/nano waste slag as a shielding panel for photons (experimental and theoretical study)

This study not only provides an innovative technique for producing rigid polyurethane foam (RPUF) composites, but it also offers a way to reuse metallurgical solid waste. Rigid polyurethane (RPUF) composite samples have been prepared with different proportions of iron slag as additives, with a range of 0–25% mass by weight. The process of grinding iron slag microparticles into iron slag nanoparticles powder was accomplished with the use of a high-energy ball mill. The synthesized samples have been characterized using Fourier Transform Infrared Spectroscopy, and Scanning Electron Microscope. Then, their radiation shielding properties were measured by using A hyper-pure germanium detector using point sources 241Am, 133 BA, 152 EU, 137Cs, and 60Co, with an energy range of 0.059–1.408 MeV. Then using Fluka simulation code to validate the results in the energy range of photon energies of 0.0001–100 MeV. The linear attenuation coefficient, mass attenuation coefficient, mean free path, half-value layer and tenth-value layer, were calculated to determine the radiation shielding characteristics of the composite samples. The calculated values are in good agreement with the calculated values. The results of this study showed that the gamma-ray and neutron attenuation parameters of the studied polyurethane composite samples have improved. Moreover, the effect of iron slag not only increases the gamma-ray attenuation shielding properties but also enhances compressive strength and the thermal stability. Which encourages us to use polyurethane iron-slag composite foam in sandwich panel manufacturing as walls to provide protection from radiation and also heat insulation.

X-ray photoelectron spectroscopy, structural, and radiation shielding properties for transparent borosilicate glasses

The current work investigates the impact of SrO content on the structural, physical, and shielding properties of xSrO-10TiO2–10SiO2-(65-x) B2O3, where (15 ≤ x ≤ 30) glasses. X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Fourier transform infrared (FTIR), and radiation shielding features were utilized to investigate the samples’ features. XRD results affirmed the amorphous behavior. The FTIR values presented many peaks attributed to SiO2 and B2O3. The Sr30B50 sample (3.048 g/cm3) showed the highest density, followed by Sr25B55 (2.936 g/cm3), Sr20B60 (2.720 g/cm3) and Sr15B65 (2.679 g/cm3). The XPS analysis of Ti 2p core level spectra reveals that Ti ions exist in both Ti4+ and Ti3+ oxidation states, with the content of Ti4+ increasing with the SrO ratio in the glass sample. The XPS analysis of O 1s core level spectra revealed that the number of non-bridging oxygen atoms is enhanced with an increase in SrO content, indicating the weakening of the glass structure. Furthermore, neutron and photon shielding properties were calculated theoretically using Phy-X software. The gamma shielding properties were enhanced with increasing SrO, while the neutron attenuation was equivalent to water. For example, the fast neutron removal cross-section (FNRCS) values are 0.101, 0.098, 0.102, and 0.101 cm−1 for Sr15B65, Sr20B60, Sr25B55, and Sr30B50, respectively. The findings suggest that the SrO improved radiation shielding properties while decreasing glass stability.

Development of HDPE/natural fiberboard nanocomposite reinforced Bi2O3/BaO and H3BO3: A sustainable approach for gamma and neutron shielding application

AbstractRadiation shielding materials significantly contribute to various fields. In this research, a new eco‐friendly nanocomposite shielding material based on high‐density polyethylene (HDPE) and natural fiberboard (NF) with different concentrations of barium oxide nanowires (BaO), bismuth oxide nanorods (Bi2O3), and boric acid (H3BO3) is constructed and investigated for 137Cs, 60Co gamma ray, and 142Am–Be neutron sources. Mechanical and physical testing and thermal analysis were conducted to assess the impact of nanostructure incorporation on the structural integrity and thermal stability of the nanocomposite. The 60Co gamma and neutron attenuation properties were evaluated using the appropriate radiation sources. The transmission and absorption coefficients of the nanocomposites were measured. The results demonstrated that HDPE/NF filled with BaO, Bi2O3 nanostructure, and boric acid has higher tensile strength and thermal stability than unfilled HDPE/NF. The results displayed that the HDPE/NF filled with 1.5% Bi2O3 and 1.5% BaO improved gamma attenuation and also HDPE/NF/10% boric acid had a high value. The outcomes of this research provide valuable insights into the design and fabrication of radiation shielding materials by understanding the relationship between nanoparticle concentration, radiation attenuation properties, and the mechanical and thermal characteristics of the nanocomposite. This study aims to optimize the performance and applicability of HDPE/NF nanocomposites for radiation protection applications.Highlights Synthesis of BaO nanowires and bismuth oxide (Bi2O3) nanorods by hydrothermal simple method. Fabrication and characterization of a new HDPE/natural fiberboard (NF) nanocomposite containing BaO/Bi2O3/H3BO3 as an eco‐friendly additive. The HDPE/NF/1.5% BaO/1.5% Bi2O3 nanocomposite has better tensile strength and more thermal stability. Usage of HDPE/NF/1.5% BaO/1.5% Bi2O3 nanocomposite that should be utilized for radiation shielding. Usage of HDPE/NF/1.5% BaO/1.5% Bi2O3/10% H3BO3 nanocomposite that should be utilized for neutron attenuation performance.

A novel aloe vera mediated Bismuth–Magnesium–Copper nanocomposite for radiation shielding applications

In the current study, Bismuth-Magnesium-Copper nanocomposites (BMC NCs) are synthesized using versatile solution combustion synthesis using Alovera extract as a reducing agent. As obtained BMC NCs are characterized using standard techniques. Bragg’s reflections confirmed the formation of multi-phase BMC NCs with crystal sizes 18.11 nm and 21.12 nm. The surface morphology of BMC NCs confirmed the agglomeration of spherical-shaped nanoparticles. EDAX spectrum analyzed the presence of Bi, Mg, and Cu. The direct energy gap is found to be 3.08 eV. X-ray/gamma radiation shielding properties are then evaluated using a NaI(Tl) detector coupled with MCA. Shielding parameters are slightly higher than lead. Neutron shielding parameters are compared with conventional shielding materials. Scattering length is found to be 8.8 fm, and neutron absorption cross-section and neutron attenuation parameters of BMC NCs are found to be efficient when compared with lead, concrete, and steel. Bremsstrahlung radiation shielding parameters are computed at 1.511 MeV. Bremsstrahlung efficiency is comparable to lead, bremsstrahlung dose rate and specific bremsstrahlung constant are found to be more in comparison with lead. Thus, BMC NCs are highly efficient in shielding X/gamma ray and neutron radiations.

Computational Codes for Fast Neutrons and Gamma Ray Interactions Coefficients in Different Material

In this research, many materials used as shields against ionizing radiation (neutrons and gamma rays) were studied. Seven glass systems compounds and five polymeric compounds were used to test their effectiveness in attenuating fast neutrons. These compounds are (S1-S7) as shown in table (1) and polymeric compounds as shown in table (3), many attenuation coefficients for these compounds were calculated through the use of a computer program called (SAZ) written in the Python programming language, which is used to calculate a number of basic parameters based on certain semi-empirical equations. The results showed that the sample (S7) and C9H9O is the best glass system compound and polymeric compound for neutrons attenuation, because they have a largest value of ΣR. Also, a number of Eco-friendly metallic compounds as shown in table (2) were used, to test them in shielding against gamma rays. The mass attenuation coefficient was calculated through the NIST-XCOM program for a range of energies starting from 59.53 keV up to 1 GeV. These compounds differed in their ability to attenuate gamma rays, as the results showed that the best studied compound against gamma rays was the PbS compound because it contains the Lead element, which has a largest atomic number.

Effect of ZrO2 on physical and radiation shielding properties of tellurium −lead- calcium −borate (TPCB)glass system

This research investigates the impact of adding Zirconia (ZrO2) to tellurium-lead-borate glasses, focusing on how it affects their physical, optical, and radiation shielding properties, particularly against gamma-rays and neutrons. Key parameters for gamma-ray shielding (mass attenuation coefficient (µm), effective atomic numbers (Zeff), and effective electron densities (Neff)) were measured across an energy range of 0.080 to 2.614 MeV. Notably, excellent agreement (within 0.5 %) was observed between the measured and calculated µm values. For slow neutron attenuation measurements, we employed a 241Am-Be neutron source in conjunction with a BF3 detector. Fast neutron attenuation was assessed using the NXCOM software to determine the effective macroscopic removal cross-section (ΣR). The addition of Zirconia notably enhanced the glass density (∼6.8 %) and its ability to shield against gamma-rays (∼10 %) and slow neutrons (∼71 %). However, this improvement comes with a slight trade-off: a decrease of approximately 1.5 % in fast neutron attenuation was observed. This trade-off between shielding effectiveness for different radiation types warrants further investigation to optimize glass compositions, considering the desired balance between shielding performance across various radiation spectrums.

Carbon analysis of large soil samples using the tagged neutron method: Accounting for radiation attenuation

Non-destructive methodology for determining carbon content in large or semi-infinite (soil) samples is discussed. This methodology is based on deconvoluting the sample's gamma spectra (received by tagged neutron method) on the sample component's spectra by accounting for neutron and gamma radiation attenuations. This algorithm was tested with both Monte-Carlo simulations and experimental gamma spectra. Good agreement was found between defined and actual sample component content. Application of this method for soil carbon determinations in agricultural fields is discussed.

Developed concrete reinforced by waste of lead fishing weights and steel nails as gamma rays and neutrons shields for nuclear power reactors

Wastes of steel nails WSN and wastes of lead fishing weights WLFW were successfully recycled in concrete to enhance its ability to attenuate neutrons and gamma rays. A series of studies on the effect of partially replacing coarse aggregates of concrete with different percentages (5, 10, and 20%) of WSN or WLFW on the density, slump, compressive strength, and attenuation of neutrons and gamma rays were implemented. The compressive strength of WSN-reinforced concrete ranged between 36.2 and 45.2 MPa, while that of WLFW-reinforced concrete ranged between 28.6 and 45.1 MPa. The ability of WSN- or WLFW-reinforced concrete to attenuate neutrons and gamma rays was tested by using two types of neutron energies; slow and total slow neutrons; and three gamma rays energies 661.64, 1173.23, and 1332.51 keV. Slow and total slow neutrons macroscopic cross-section, linear and mass attenuation coefficients of gamma rays, and half value layer for both types of nuclear radiation were obtained. The results showed that there is a significant improvement in the slow and total slow neutron attenuation efficiency of the considered concrete when reinforced by WSN or WLFW. On the other hand, due to the increase in the atomic number and density of both Fe present in WSN and Pb in WLFW, the attenuation ability of the studied gamma ray energies increased. The WLFW-reinforced concrete showed better value in the gamma rays and neutrons attenuation parameters compared to the WSN-reinforced concrete. Hence, the considered concretes reinforced by WSN or WLFW are a distinguished choice as radiation shields in nuclear power reactors, in addition to their adaptability to high concentrations of waste.

Thermoplastic and thermoset polymer matrix composites reinforced with bismuth oxide as radiation shielding materials

Polymer composites reinforced with heavy metals have wide applications in the nuclear industry as radiation shielding materials against emitted radiation from nuclear equipment. In this study, considering bismuth oxide as reinforcement, various polymers including thermoplastic polymers, such as high-density polyethylene, low-density polyethylene, polypropylene, polyvinyl acetate, polymethyl methacrylate, and thermosetting epoxy resin have been considered as matrices. After preparing various samples of polymer composites loaded with 60 % by weight of bismuth oxide, some physical, mechanical, and radiation shielding properties (neutron and gamma attenuation coefficients) of these polymer composites have been measured and compared with each other. First, the density of the samples was measured, and the response of these samples to varying compressive force was investigated. The stress-strain diagrams of composites with different matrices were compared. Subsequently, the linear and mass attenuation coefficients of the composites were measured over a wide range of energies (32–1480 keV) against direct and collimated beams from various standard gamma-ray sources (134Cs, 60Co, 154Eu, 133Ba, and 22Na) using an HPGe detector. A collimated beam from an Am–Be source was utilized to measure and compare the attenuation properties of the composite samples against neutron flux. Furthermore, the properties of the composites against neutron and gamma radiation were calculated using the MCNP Monte Carlo code over a broad energy range, and the results were compared and validated with the experimental data. The comparison of the calculated and measured results shows reasonable agreement. The results indicate that in situations where high attenuation properties coupled with high strength are desired, polymeric composites with an epoxy matrix can be considered as radiation shielding materials. If the use of flexible and malleable material is required, such as for reactor piping shielding or the fabrication of protective gloves and clothing in a radiation environment, among the polymers studied, polyvinyl acetate is a more suitable matrix.

Microstructure and radiation shielding properties of lead-fiber reinforced high-performance concrete

The increasing application of nuclear technology in medical fields has amplified the demand for radiation shielding materials. Traditional large-volume concrete cannot simultaneously exhibit excellent neutron and gamma-ray shielding capabilities while ensuring superior mechanical properties. This study tailored a novel lead fiber-reinforced high-performance concrete (LFRHPC) with lead fibers and magnetite aggregates, focusing on radiation attenuation. The effect of lead fibers on the mechanical performance, microstructure and radiation attenuation characteristics for LFRHPC was comprehensively evaluated. The results indicate a significant improvement in the mechanical performance of LFRHPC compared to traditional radiation shielding concrete (RSC), due to the scientific skeleton design. Additionally, the pore structure in LFRHPC is refined by incorporating lead fibers. Moreover, the interfacial transition zone (ITZ) between the cementitious matrix and magnetite is improved because of the very low water-to-binder ratio. To assess the radiation attenuation ability of LFRHPC, both direct measurements and simulation models were utilized. The data reveals that the porosity of LFRHPC plays a crucial role in its shielding effectiveness against neutron and gamma radiation. Furthermore, the addition of lead fibers considerably boosts its neutron and gamma-ray shielding attenuation capabilities. In summation, LFRHPC combines excellent mechanical properties and radiation shielding capabilities, making it suitable for construction applications in medical settings.

Synthesis and characterization of Bi2O3–B2O3–SiO2–MgO glasses: Structural, optical, and shielding properties

The systematic investigation of the impact of Bi2O3 on the structural, spectroscopic, and radiation features of 15SiO2 - 75B2O3 - (10-x) MgO - xBi2O3, x= (0 ≤ x ≥ 10) glasses were studied. The increase in glass density as Bi2O3 content increase is due to the incorporation of Bi2O3 into the glass network. XRD patterns indicate an amorphous, glassy nature of the glasses. The optical absorption spectra suggest that as the content of Bi2O3 increases, the optical band gap (Eopt.) widens and the Urbach energy (Eu) decreases. It has been observed that the opposite trend between the (Eopt.) and (Eu). The radiation shielding characteristics of MgBi glasses have been evaluated through the assessment of several fundamental radiation attenuation variables. As Bi2O3 content increases would enhance properties like (HVL), (MFP), and (Zeq), especially at lower energy ranges. Adding Bi2O3 to the glassy system has a positive impact on enhancing its neutron attenuation ability. The MgBi-10 glass has the best shielding performance against fast neutrons. The MgBi-10 glass sample shows promising properties for optical and radiation shielding purposes.