MCNP5 Simulation Code Research Articles

Investigation of mechanical and radiation shielding properties of Sm–Sm2O3 reinforced Al–B4C composite

In this study, Al–20%B4C-x%(Sm, Sm2O3) (x = 1, 3, and 5) composites were prepared using mechanical alloying to investigate their neutron shielding and mechanical properties. After pelletizing and sintering the prepared powder composites, their microstructures were characterized using X-ray diffraction, energy-dispersive X-ray spectroscopy, and scanning electron microscopy. Hardness, density, and corrosion tests were performed on the composites. Thermal and fast neutron shielding properties were calculated using the MCNP simulation code. Experimental measurements of neutron shielding were carried out using an Am–Be neutron source. Experimental results showed that 5% Sm-reinforced composites exhibited better neutron shielding than Sm2O3-reinforced composites at all doping ratios. Simulation calculations indicated that the highest thermal neutron shielding rate was observed in Sm-reinforced composites, while the highest fast neutron shielding rate was observed in 5% Sm2O3-reinforced composites. This research provides important data for researchers interested in developing thermal and fast neutron shielding materials.

A good balance between the efficiency of ionizing radiation shielding and mechanical performance of various tin-based alloys: Comparative analysis

The Sn-Bix (x = 30, 40, 50, 60 and 70%) radiation shielding alloys were prepared using a melting process of the alloying material in a ceramic crucible at 200 °C for 120 min. The structural characterization for synthesized samples was carried out using X-ray Diffraction technique (XRD). The hardness and elastic properties of the prepared alloys was measured using micro-indentation creep technology and tensile test machine respectively. XRD pattern indicated that that the crystal size of Sn phase is lowered by increasing Bi concentration in the Sn matrix. The hardness decreases as the dwell time increases. The hardness values and elastic properties (in terms of young modulus) were significantly improved due to the decrease in the crystallite size by increasing Bi content. The stress exponent (n) value increase by increasing Bi content which mean that the mechanical properties improved due to increment of resistance. Furthermore, the produced alloys' nuclear shielding ability was investigated. The μm values were calculated using MCNP simulation code and XCOM software over a broad energy range of 0.015 MeV–15 MeV with good agreement between them. Several characteristics are estimated in order to properly understand the researched alloy's radiation and properties of neutron shielding. The findings showed that a higher Bi concentration causes the crystal size to decrease, improving the radiation shielding and mechanical qualities. The alloy Sn-70% Bi has a maximum increase of vicker hardness, tensile strength, and young's modulus. The findings of the shielding parameters indicate that the alloys under study are efficient gamma shielding materials in contrast to other typical shielding materials and materials that have been examined recently. The highest mechanical efficiency and rather strong gamma-ray shielding capabilities are found in the Sn–70Bi alloy. Owing to its ability to effectively balance mechanical performance and shielding, the Sn–70Bi are approved for use in radiation protection. In addition, when compared to all other manufactured alloys, typical neutron shielding materials, and recently investigated materials, the results further demonstrate that the Sn–70Bi alloy has the best neutron attenuation capability. Furthermore, in terms of mass stopping power (MSP) and projected range (PR), the Sn–70Bi alloy demonstrates the most effective attenuation for alpha particles (He+2) and protons (H+1). These results imply that the Sn–70Bi alloy provide superior mechanical performance and nuclear shielding for a range of applications including industrial, medicinal, and nuclear waste storage in the future.

Influence of WO3 replacement for CaO on physical, optical, and γ-ray protection properties of borotellurite glasses: A comparative study

Glasses with WO3 in place of CaO with the chemical formula 45B2O3+15TeO2+(18-X)CaO+22Na2O + XWO3: X = 0 (W-0), 1.5 (W-1.5), 3 (W-3), 4.5 (W-4.5), and 6 (W-6) mol% were prepared using the melt quenching method. A comprehensive study of the structural, physical, optical, and γ-ray protection properties of the prepared glasses was investigated. Density (ρ), molar volume (Vm), UV–Vis measurements, the MCNP5 simulation code, and PhX software were employed to achieve the aims of this study. The density (ρ) of the W-X samples increased slightly from 2.45 g/cm3 to 2.62 g/cm3 as WO3 content increased from 0 to 6 mol%. The molar volume (Vm) evinced the same trend as ρ. In the UV range, significant absorption was observed, rising as the molar concentration of WO3 increased. The direct optical gap (Eopt.(dir.)) declined from 3.59 eV to 3.24 eV, while the indirect gap (Eopt.(indir.)) declined from 3.27 eV to 2.36 eV. The Urbach energies (ΔE) ranged from 0.35 to 0.71 eV. The refractive index (n) increased from 2.33 to 2.59 as WO3 content increased. The molar polarizability (αm) and molar refractivity (Rm) increased. The linear-attenuation absorption (μ) order is: W-0 < W-1.5 < W-3 < W-4.5 < W-6. The W-6 glass sample possessed the lowest half- (HVL), and tenth- (TVL) value layer, as well as the lowest mean free path (MFP). Therefore, the W-6 glass sample offers the best gamma radiation shielding capability among the prepared W-X glasses. Within the investigated energy range, the effective atomic number (Zef) varied from 30.773 to 13.234, 35.398–13.969, 39.089–14.670, 42.103–15.338, and 44.611–15.977 for the W-0, W-1.5, W-3, W-4.5, and W-6 glasses, respectively. At 0.080 MeV, the mass-attenuation absorption (μm) of the investigated W-X glasses was higher than for TBLMC-X (TeO2–B2O3–Li2O–MoO3–CuO) and TS-X (9SiO2–Al2O3–Na2O–B2O3–TeO2) glasses.

Optical and radiation shielding properties of PVC/BiVO4 nanocomposite

This study investigates the physical and optical properties as well as the radiation shielding capacity of polyvinyl chloride (PVC) loaded with x% of bismuth vanadate (BiVO4) (x = 0, 1, 3, and 6 wt%). As a non-toxic nanofiller, the designed materials are low-cost, flexible, and lightweight plastic to replace traditional lead, which is toxic and dense. XRD patterns and FTIR spectra demonstrated a successful fabrication and complexation of nanocomposite films. In addition, the particle size, morphology, and elemental composition of the BiVO4 nanofiller were demonstrated through the utilization of TEM, SEM, and EDX spectra. The MCNP5 simulation code assessed the gamma-ray shielding effectiveness of four PVC + x% BiVO4 nanocomposites. The obtained mass attenuation coefficient data of the developed nanocomposites were comparable to the theoretical calculation performed with Phy-X/PSD software. Moreover, the initial stage in the computation of various shielding parameters, such as half-value layer, tenth value layer, and mean free path, besides the simulation of linear attenuation coefficient. The transmission factor declines while radiation protection efficiency increases with an increase in the proportion of BiVO4 nanofiller. Further, the current investigation seeks to evaluate the thickness equivalent (Xeq), effective atomic number (Zeff), and effective electron density (Neff) values as a function of the concentration of BiVO4 in a PVC matrix. The results obtained from the parameters indicate that incorporating BiVO4 into PVC can be an effective strategy for developing sustainable and lead-free polymer nanocomposites, with potential uses in radiation shielding applications.

Synthesis, optical, mechanical characteristics, and gamma-ray shielding capacity of polyethylene -basalt mixture

In this work, a new polymeric composite consisting of polyethylene polymers as a base material filled with different weight percentages of basalt rock was prepared. The basalt ratio was added to the base material with percentages of 1.0 up to 20.0 wt%. The crystal structure, optical, mechanical properties, and gamma-ray shielding capacity of the undoped and the doped polymeric samples which were coded as HDPE-Bx, where x = 0.0, 1.0, 3.0, 5.0, 10.0, and 20.0 wt% were investigated. The optical energy band gap (EOptical) of the prepared mixture was decreased by increasing the basalt concentration in the mixture from 4 to 2.6 eV for the allowed direct transition and from 5.5 to 4.7 eV for the allowed indirect transition. The mechanical properties of the HDPE polymers were enhanced with the enhancement of Basalt (powder) added to the host polymers with ratios ranging from 0.0 to 20.0 wt%. Radiation shielding capacity of the prepared samples was investigated in the 0.015–15 MeV photon energy via Phy-X software and MCNP-5 simulation code. Results revealed that the mass attenuation coefficient (μm) of the prepared samples increased with increasing the basalt ratio from 1.0 to 20 wt%. Besides, the half-value layer (HVL) decreased. The μm ranged from 0.7455 to 2.956 cm2/g for HDPE-B0 to HDPE-B20 at 0.015 MeV photon energy, while at 15 MeV, the result refers to an increase in the μm from 0.0181 to 0.0189 cm2/g for HDPE-B0 to HDPE-B20. The HVL of the prepared samples was arranged in the order that HDPE-B0 > HDPE-B1 > HDPE-B3 > HDPE-B5 > HDPE-B10 > HDPE-B20. This order assured that HDPE-B20 is better in shielding applications than the other prepared samples.

Study of the effect of photon self-absorption on the detection efficiency of HPGe detector in measurements of an activated uranium sample

The photon self-absorption effect of uranium on the full-energy peak (FEP) efficiency calibration of high purity germanium (HPGe) detector was studied as a function of energy using an activated uranium sample has a thickness of 1 mm. The absolute efficiencies of the detector were obtained with the use of gamma standard sources and gamma rays of radionuclides created by irradiating the uranium sample with secondary neutrons. The measurements were carried out with the use of a Planar HPGe detector. MCNP simulation code was used to verify the experimental results. The comparison of the empirical results with the simulation results showed a good agreement. To interpret the efficiencies, the reaction rates were calculated using two methods: (1) - the use of the efficiency curve from the standard sources combined with the self-absorption correction of the uranium material and (2) - the use of the efficiency curve obtained from gamma rays in the sample.

Bismuth oxide nanoparticles (Bi2O3 NPs) embedded into recycled- Poly(vinyl chloride) plastic sheets as a promising shielding material for gamma radiation

In this work, novel γ-ray shielding materials based on recycled poly (vinyl chloride) (r-PVC) were prepared via melt mixing and compression molding technique. The designed materials are low-cost plastic wastes, ductile and lightweight, doped with xBi2O3 NPs (x = 0.0%, 5.0%, 15.0%, 25.0% and 35.0% (w/w)) as a non-toxic nanofiller. Scanning electron microscopy (SEM) and X-ray diffraction (XRD) were used to explore the characteristics of the as-prepared r-PVC/Bi2O3 nanocomposites. The γ-ray shielding effectiveness of the r-PVC/Bi2O3 nanocomposites were evaluated using NaI(Tl) scintillation detector at different γ-ray photon energies (121, 334, 778, 963, 1112 and 1408 keV) emitted from Eu-152 standard radioactive point source. For each of the mentioned photon energies, the mass attenuation coefficients (μm) were measured and compared to the Bi2O3 NPs’ doping ratio. The results showed that as the content of Bi2O3 NPs was increased in the nanocomposite, the value of μm significantly increased. In particular, the r-PVC nanocomposite sample containing 35.0 wt% Bi2O3 NPs exhibited the highest μm (2.5301 ± 0.22 at 121 keV). The μm was also calculated theoretically, using the MCNP5 simulation code. Experimental data of μm for all the tested nanocomposites were comparable to that obtained from a theoretical calculation. Moreover, the experimentally measured linear attenuation coefficient was the first step in calculating other shielding parameters, including half-value layer (HVL), tenth value layer (TVL), and mean free path (MFP). These parameters confirmed that addition of Bi2O3 NPs resulted in a higher shielding efficiency for r-PVC. Finally, this study shows that addition of Bi2O3 NPs to r-PVC facilitates the production of lead-free and sustainable polymer nanocomposites with improved γ-ray shielding characteristics, which can be used in nuclear energy applications, as alternatives to the shielding materials currently in use.

Suggested two layers container for shielding the low and intermediate activity gamma-ray sources

Due to the advantages of using gamma-ray radiation sources in various fields, the average equivalent dose received due to the exposure to these sources increases. The wrong handling or direct exposure to these sources give raises the equivalent dose for the receiver. Thus, a container made up of two layers (metal and glass layers) was suggested to decrease the recorded equivalent dose received from the gamma-ray radioactive sources to an acceptable limit. The Monte Carlo MCNP simulation code was utilized to predict the equivalent dose around the container's outer surface and at 100 cm from the center of the outer container's faces. For example, the simulation study illustrated that the equivalent dose rate received at 100 cm from an unprotected 137Cs source with an activity of 3.7E+8 Bq is 26.30 μSv/h. This equivalent dose rate from an unprotected source is reduced to 22.75 μSv/h using the inner layer of the stainless steel with a thickness of 0.5 cm. After that, the equivalent dose rate reduced from 22.75 to 7.60 μSv/h, raising the WO3. Simultaneously, the equivalent dose rate from an unprotected Co-60 source reaches 50.51 μSv/h. It is reduced to 23.56 μSv/h by using a container consisting of an inner layer of stainless steel with a thickness of 0.5 cm and an outer glass layer of BaWP7 glass sample with WO3 content of 60 mol % and thickness of 5 cm.The present study showed the suggested container's ability in reducing the absorbed dose and equivalent dose rate for intermediate activity gamma-ray sources used in laboratories and nuclear medicine.

Impact of B2O3 on physical, optical characteristics and radiation attenuation factors of borotellurite glasses

Boro-tellurite glasses with a composition [(60-X)TeO2-(20 + X)B2O3–10Li2O–10Bi2O3] where X = 5, 10, 15, 20 mol% have been synthesized. Glass density, molar volume, oxygen packing density, and many other physical parameters were measured. Ultraviolet (UV)-spectra in the wavelength range of 200–800 nm have been measured for the whole glass series. The optical energy band gap Eopt, refractive index, and optical basicity were measured. The mass absorption coefficients (μm) are determined experimentally by the HPGe detector and compared with the theoretical values obtained by the XCOM program and MCNP5 simulation code within 0.121–1.408 MeV photon energy range. Half value layer (HVL), effective atomic number and electron density (Zeff and Neff), and macroscopic removal cross-section (∑R) were evaluated. The sample [55TeO2 – 25B2O3 – 10Bi2O3 – 10Li2O] possess the highest values of (μm = 1.192 ± 0.033 cm2/g, Zeff = 56.12 e/atom and ∑ R = 0.101499 cm−1) at energy 121 keV also lower values of (HVL = 0.121 cm, TVL = 0.1 cm and MFP = 0.174 cm) at photon energy 121 keV. Therefore [55TeO2 – 25B2O3 – 10Bi2O3 – 10Li2O] glass sample is considered the best gamma-ray shielding material among the prepared glasses.

INVESTIGATING THE CHARACTERISTICS OF LARGE-VOLUME PVT SCINTILLATION DETECTORS IN THE RADIATION PORTAL MONITORS USING MONTE CARLO SIMULATIONS

This paper presents the characteristics determination process of the large-size Polyvinyl Toluene (PVT) scintillation detectors using MCNP5 simulation code. The energy spectra using a 137Cs calibration source, absolute efficiency in the energy range of 50 ÷ 3000 keV, and the angular response of the EJ-200 50×50×5 cm3 and 25×25×5cm3 are investigated. The simulated energy spectra are in good agreement with the experimental spectra. The results of determining the absolute efficiency show that the EJ-200 50×50×5 cm3 and 25×25×5cm3 plastic detectors have detection efficiencies of 16,3% and 9,2%, respectively, at 10cm source-to-detector distance, and down to 0,6% and 0,17% at 100 cm source-to-detector distance. The angular responses of the detectors show that the detection efficiency value reached ≥ 90% of the maximum value with the incident angle less than 5π/6. The results can be applied in the process of design optimization of plastic-based radiation portal monitors.

Influence of ZnO to the physical, elastic and gamma radiation shielding properties of the tellurite glass system using MCNP-5 simulation code

The main goals of this current work are to study the elastic and radiation shielding properties of seven different compositions of ZnO-TeO2 transparent glass and characterize them using density, molar volume, X-ray diffraction (XRD), ultrasonic technique, and Monte Carlo simulation method. From the ultrasonic wave measurement result, the substitution of ZnO, which works as a network modifier into the TeO2 glass systems, would break up the Te-O-Te bonds of TeO4 into the form TeO3 along with all the formation of NBO's which give the impact on the elastic moduli analysis. The radiation shielding features for the fabricated samples were studied using the MCNP-5 code. The difference between the MCNP and XCOM results was found to be less than 3%. The highest linear attenuation coefficient (LAC) for all ZT glasses achieved at 0.015 MeV and varied between 202.199 and 239.607 cm−1 for glasses coded 0 ZT and 30ZT, respectively.

Influence of PbO content on the gamma ray shielding properties of lead boro-telluro-phosphate glasses

This paper investigates the influence of lead monoxide (PbO) content on the gamma ray attenuation coefficients of boro-telluro-phosphate glasses. The MCNP5 simulation code has been applied to estimate the mass attenuation coefficients (μ/ρ) of Sm3+ doped (60-x)B2O3-xPbO-14TeO2-10P2O5-7.5BaO-7.5CdO-1Sm2O3 (x = 0, 10, 20, 30 and 40 wt%) glass system within 0.02–10 MeV photon energy range. The MCNP5 simulated results are accord with the theoretical XCOM results perfectly. The results show that the increasing content of PbO in the glass compositions can improve the gamma ray shielding capability of the studied glasses. With the biggest content of PbO in Pb40B20 glass, superior HVL and Zeff values are observed. The values of HVL for Pb40B20 glass vary from 0.0035 to 3.6824 cm, and the values of Zeff vary from 25.94 to 70.05 depending on the photon energy 0.02–10 MeV. In addition, the boro-telluro-phosphate glasses have been compared with types of commercial concretes and glasses in terms of HVL.

Fabrication of TeO2-doped strontium borate glasses possessing optimum physical, structural, optical and gamma ray shielding properties

Glasses with composition of (60 − x) TeO2 − 10SrO − (30 + x) B2O3 were fabricated by melt quenching method. The structural features have been studied using X-ray diffraction (XRD) and FTIR. The derivative of absorption spectra fitting (DASF) method has been utilized to determine the optical band-gap energy. The measured band-gap energies takes values 3.521, 3.558, 3.581, 3.620 and 3.669 eV for the BSrTe60, BSrTe55, BSrTe50, BSrTe45 and BSrTe40, respectively. The band-gap energies increase with a decrease in the mol% of concentration of TeO2. The transitions involved are found to be indirect transitions. The MCNP-5 simulation code was put to use in estimating gamma photons' mean path length in the material. The lowest linear attenuation coefficient (µ) values lie within the range of 0.162 and 0.107 cm−1 and are detected around 8 MeV. The LAC values decreased linearly with substituting the TeO2 by B2O3 content. The lowest half-value layer (HVL) is found at low energy (around 0.005 cm at 0.015 MeV). The mean free path (MFP) decreases linearly with increasing the TeO2 content and followed the trend: (MFP)BSrTe60 < (MFP) BSrTe55 < (MFP) BSrTe50 < (MFP) BSrTe45 < (MFP) BSrTe40. The MFP of BSrTe60 and BSrTe55 at 0.662 MeV is 2.620 and 2.950 cm, and this is lower than the MFP of some commercial glasses such as RS 253 (MFP = 5.263 cm, RS 253-G18 (MFP = 5.263 cm), RS 323-G19 (MFP = 3.571 cm) and RS 360 (MFP = 3.125 cm) at the same energy.

Gamma ray exposure buildup factor and shielding features for some binary alloys using MCNP-5 simulation code

Gamma radiation shielding features for three series of binary alloys identified as (Pb–Sn), (Pb–Zn), and (Zn–Sn) have been investigated. The mass attenuation coefficients (μ/ρ) for the selected alloys were simulated using the MCNP-5 code in the energy range between 0.01 and 15 MeV. Moreover, the (μ/ρ) values were computed using WinXCOM database in the same energy range to validate the simulation results. Results reveal a good agreement between the simulated and computed values. The half value layer (HVL), mean free path (MFP), effective atomic number (Zeff) and exposure buildup factor (EBF) were evaluated for the selected binary alloys. Results showed that the PS1, PZ1, and ZS2 alloys have the best shielding parameters and better than the commercially standard and available radiation shielding materials. Therefore, the investigated alloys can be used as effective radiation shielding materials against gamma ray with energies between 0.01 and 15 MeV.

Monte Carlo simulation of radiation shielding properties of the glass system containing Bi2O3

Monte Carlo simulation is an important tool to obtain some parameters of a material, especially when it is impossible to perform an experiment. Radiation shielding of a material has been very important properties as the radiation started to be used in a variety of different fields. Besides using it in different fields, thus protection from its hazardous effects becomes a popular subject for researchers. For this purpose, there have been many different works done for different types of materials. Glass is an important material, and it is interesting to investigate its radiation shielding properties. Thus, in the present study, waste soda–lime–silica-based glasses (SiO2–Na2O–CaO–Bi2O3) were calculated using MCNP simulation code, and the results were compared with the XCOM calculation results.

Evaluation of gamma-rays attenuation competences for waste soda-lime glass containing MoO3: Experimental study, XCOM computations, and MCNP-5 results.

The present work aims to examine a waste soda-lime-silica (SLS) glass system containing varying amounts of MoO3 for its potential use in radiation shielding applications. A series of 4 different glasses (RG, RGM1 to RGM3) were prepared. The starting materials were precisely weighed and mixed homogeneously. The glass samples were then synthesized by melting the mixtures in a Au-Pt crucible via a conventional electrical resistance furnace at 1300 C. The fabricated glasses were analyzed by performing physical and chemical measurements. Some glass property calculations were done to reveal the effect of changing the contents of MoO3. According to the results, all the fabricated SLS glass samples showed an amorphous structure with a suitable transparent appearance. The density and refractive index values increased as the amount of MoO3 in the glass systems increased. Moreover, the gamma-ray shielding ability of the SLS glass network was evaluated experimentally using a NaI (Tl) detector. The measured values of the linear attenuation coefficient (LAC) were confirmed using MCNP-5 simulation code and XCOM program. The LAC increased from 0.184 to 0.235 cm−1 when increasing the MoO3 ratio from 0 to 0.5 wt%, respectively. An agreement between the experimental measurements and theoretical calculations was achieved. The difference between the experimental and theoretical values was within 5 %. Based on the experimental LAC and other essential parameters, such as the half-value layer (∆0.5), mean free path (MFP), transmission factor (TF), and radiation protection efficiency (RPE), the radiation shielding ability of the investigated SLS glass system was determined.

Ta2O5 reinforced Bi2O3–TeO2–ZnO glasses: Fabrication, physical, structural characterization, and radiation shielding efficacy

The melt-quenching method was applied to fabricate a new glass series containing 10Bi 2 O 3 –70TeO 2 -(20-x)ZnO-xTa 2 O 5 , x = 0, 2, 4, and 6 mol%. The chemical compounds were well mixed and melted in a furnace at 900 °C, and then the molten transferred to another furnace at 320 °C for annealing. X-ray diffraction is used to confirm the amorphous phase of the fabricated glasses. Moreover, the fabricated glasses' structure characterization was established using Raman analyses. Besides, the gamma-ray protection capacity was evaluated for the fabricated glass series. Thus, the MCNP-5 simulation code and XCOM software program were utilized to estimate the investigated glass samples' linear attenuation coefficients. The glass sample's highest attenuation coefficient, with 6 mol % of the Ta 2 O 5 doping ratio is ranged between 418.17 and 0.27 cm −1 . In comparison, the fabricated glass without Ta 2 O 5 possesses the lowest attenuation factor and decreased from 359.39 cm -1 to 0.26 cm −1 , raising the incoming photon energy from 0.015 MeV to 15 MeV, respectively. The simulated linear attenuation coefficient was then applied to calculate the half-value thickness, and transmission factor for the incoming photons. The obtained results showed that the fabricated glass samples' shielding ability enhanced with the insertion of the Ta 2 O 5 . • New glass system of chemical composition 10Bi 2 O 3 -70TeO 2 -(20-x)ZnO-xTa 2 O 5 ; where x=0, 2, 4, and 6 mol% have been prepared. • Density and molar volume values were calculated for each synthesized sample by using the Archimedes principle. • Raman spectra were obtained from the region of 100-4000 cm -1 . • Shielding capacity for glass samples have been investigated via MCNP-5 and Xcom.

Investigation of Gamma‐Radiation Shielding Properties of Cadmium Bismuth Borate Glass Experimentally and by Using XCOM Program and MCNP5 Code

New glass systems of bismuth borate with various concentrations of cadmium oxide are prepared based on the melt‐quenching method. The X‐ray diffraction (XRD) reveals a fully amorphous structure of the prepared glasses (S1–S4), and the UV–vis results display good transparency (&gt;50%) in the visible and near‐UV region. In addition, the radiation shielding properties (mass attenuation coefficient, half‐value layer, tenth value layer, mean free path, effective atomic number, and electron density) of the new glass system are determined at selected energies experimentally and by using MCNP5 simulation code and XCOM computer program. Based on the calculated relative difference, the obtained values from MCNP5 and XCOM are in good agreement with the experimental data. The mean free path of the current systems (particularly S4) shows optimistic results when compared with the barite and chalcocite concretes.

Investigating the effect of basalt fiber additive on the performance of clay barriers for radioactive waste disposals

Radioactive materials are widely used in mining, manufacturing, medicine, and agricultural processes. The waste products from these materials are hazardous and should be appropriately dumped. As a result, a proper radiation shielding barrier is required to avoid the contamination of the surrounding environment. Clay soil is an efficient and eco-friendly radiation shielding material, which is commonly used to cap the hazardous and radioactive landfills. In this research, the effects of basalt fiber additive in four percentages, including 0.5, 1, 2, and 5 on the bentonite clay radiation shielding performance, were investigated using experimental and simulation methods. Also, the permeability of the mixtures is controlled to be in the acceptable range as a vital parameter for radioactive disposal barriers. Chemical and microstructure analyses were conducted on the utilized material, using energy-dispersive X-ray spectroscopy (EDX) and scanning electron microscopy (SEM). Linear attenuation coefficient (μ), representing the material radiation shielding performance, was evaluated using the HPGe spectrometer detector, MCNP simulation code, and XCOM web program on the three commonly used gamma-ray energy levels of 137Cs (661.6 keV) and 60Co (1173.5 and 1332.5 keV). An acceptable agreement was observed between experimental and simulation results, revealing that adding basalt fiber improves the shielding performance due to a higher linear attenuation coefficient (μ), where 2% of basalt fiber leads to the highest values of 12.3 m−1, 10.14 m−1, and 8.5 m−1 obtained for 661.6 keV, 1173.2 keV, and 1332.5 keV energy levels, respectively. The results concluded that, due to radiation shielding performance, workability, and permeability limitations, the 2% basalt-bentonite mixture could be a new candidate covering low-level radioactive waste disposal.

Tm3+ ions-doped phosphate glasses: nuclear shielding competence and elastic moduli

Photon and neutron shielding factors and elastic moduli of phosphate glass doped with Tm3+ ions (abbreviated as PPKANT glasses) have been analyzed. MCNP-5 simulation code, and Phy-X/PSD software have been utilized to estimate the gamma rays and neutron shielding capacity. Furthermore, Makishima and Mackenzie’s theory has been performed to calculate elastic moduli. Young’s modulus, bulk modulus, and shear modulus have been varied from 40.71 to 43.62 GPa, 28.01 to 30.65 GPa, and 16.18 to 17.27 GPa, respectively. Poisson’s ratio changed from 0.257 for PPKANT0.1 glass to 0.263 for PPKANT4 glass. At low gamma photon energy (0.015 meV), the LAC increased with the increase of Tm3+ content (LAC = 155.54 and 170.18 cm−1 for glass samples PPKANT0.1 and PPKANT4, respectively). At high energy (15 MeV), the LAC changed from 0.118 to 0.124 cm−1 for PPKANT0.1 to PPKANT4. The highest values of the HVL decreased from 5.734 to 5.570 cm. The trend of MFP was like the HVL trend. At 15 MeV, the values of the EBF and EABF were rapidly increased with the increase in the penetration depth (PD), especially for PD > 20 mfp. The lowest ∑R was achieved for glasses PPKANT4 with 4 mol % Tm2O3, where ∑R = 0.02561 cm2g−1. The investigated PPKANT glasses can be considered good neutron and photon shields compared to some traditional radiation shielding materials.