Effective Removal Cross-section Research Articles

MRCsC: A user-friendly software for predicting shielding effectiveness against fast neutrons

A new user-friendly software called MRCsC was developed to accurately and precisely predict macroscopic effective removal cross-section, ΣR, (in cm−1) of fast neutrons for different shielding materials. The program includes a revised calculation model and the latest data published by Evaluated Nuclear Data Library “ENDF/B-VIII”. Experimental data and NXCom results were used to comparatively and statistically assess MRCsC predictions. The results reveal that both MRCsC and NXCom programs provide accurate and precise results when compared with the experimental data and that only MRCsC results are statistically equal to the experimental ones to the 95% confidence.

The effect of incorporated recycled low density polyethylene (LDPE) on the fast neutron shielding behaviour (FNSB) of clay matrix using MCNP and PHITS Monte Carlo codes

In order to develop a cheaper, environmentally friendly and self-sustaining high hydrogen content material for fast neutron shielding, clay-polyethylene composites containing different weight fractions of recycled low density polyethylene (LDPE) were fabricated. The effect of the incorporated LDPE on the fast neutron shielding behaviour (FNSB) of the clay matrix was investigated using MCNP and PHITS Monte Carlo codes. This was done by evaluating the fast neutron effective removal cross section (ƩR), mean free path (λ) and the neutron yield of the fabricated composites. The obtained values of effective removal cross section ƩR for the composites range between 0.056 cm−1 and 0.079 cm−1, while the values of mean free path λ are between 17.857 cm and 12.658 cm. The results show that the effective removal cross section of the samples significantly increased, while the mean free path decreased as the weight fraction of recycled LDPE increased. This implies that the fast neutron shielding behaviour of the composites improves as the LDPE concentration increases in the clay matrix. The neutron yield of the composites obtained PHITS code follows similar trend.

Enhancement of the Shielding Capability of Soda–Lime Glasses with Sb2O3 Dopant: A Potential Material for Radiation Safety in Nuclear Installations

Elastic moduli were theoretically computed using the Makishima–Mackenzie model for SiO2–Na2O–CaO glasses doped with Sb2O3 contents. The calculated elastic moduli (Young’s, bulk, shear, and longitudinal modulus) were observed to increase with an increase in the Sb2O3 contents. The microhardness showed an increase, while Poisson’s ratio decreased with the rise of the Sb2O3 contents. In addition, gamma-ray and neutron shielding parameters were evaluated for the investigated glasses. The linear attenuation coefficient (LAC) was simulated using the Monte Carlo N-particle transport code (MCNP-5). Other parameters, such as the mass attenuation coefficient (MAC), transmission factor (TF), and half-value layer, were calculated based on the simulated LAC. The addition of Sb2O3 content was observed to enhance the investigated glasses’ shielding parameters, where the highest LAC was achieved for the SCNSb10 glass with 10 mol% Sb2O3 and decreased from 0.441 to 0.154 cm−1 at gamma energies between 0.248 and 1.406 MeV. Furthermore, the fast neutron effective removal cross-section (∑R) was computed theoretically. The calculated results showed that the highest ∑R was equal to 0.0341 cm2g−1 and was obtained for the SCNSb0 glass, which had no Sb2O3 content, while the lowest ∑R was equal to 0.0286 cm2 g−1 for the SCNSb10 glass sample. The present work was carried out to examine the advantages of the soda–lime glasses with different Sb2O3 contents in several photon shielding applications, especially for radiation safety in nuclear installations.

Evaluation of physical and radiation shielding properties of Bi2O3–B2O3 glass doped transition metals ions

Bismuth borate glasses of formula 60% Bi2O3 and 40% B2O3 doped 0.7 wt % of one of the transition metal oxides; TiO2, Cr2O3, V2O5 or Fe2O3 were prepared using melting-annealing method. The glasses optical transmittance spectra have been studied and reveal UV transmission bands related to the different valence states of the doped transition metal ions. Electrical conductivity and many physical parameters of the glasses were also investigated such as; optical bandgap (Eopt), refractive index (n), reflection loss (RL), molar refractivity (Rm), molar polarizability (αm), dielectric constants (ε), optical transmission coefficient (T) and metallization criterion (M Criterion).The effectiveness of the glasses as radiation protective shield was evaluated and calculated by Monte Carlo N-Particle Transport Code (MCNP) using the Medical Internal Radiation Dose (MIRD) of male human model. Many shielding parameters have been also deliberated such as; linear attenuation coefficient, effective removal cross section (ΣR), radiation protective efficiency (RPE %), tissue weighting factors (Wt), in addition to a comparison between the experimental and the simulated linear gamma attenuation coefficients at numerous energies. The glasses reveal high stable optical and physical properties in addition to high radiation shielding efficiency against gamma irradiation. The high concentration of heavy Bi3+ ions and presence of transition metal ions with their variable configurations, provide the shielding behavior of the glasses and decreased the life time risk for cancer by about five times.

Assessment of a polymeric composite as a radiation attenuator and a restoration mortar for cracking in biological shields

This work is dedicated to figuring out robust epoxy/magnetite/boron carbide (EP/Mag/B4C) composite for radiation attenuation at multiple applications related to nuclear installations, as well as restoration mortar for cracking developed in concrete biological shields. The mechanical properties (flexural, compressive, and impact strengths) and the physical properties (water absorption, porosity, and dry bulk density), each, have been performed to label the composite integrity for practical application. In practice, attenuation properties have been performed by using a collimated beam emitted from spontaneous fission 252Cf (100 μg) neutron source and neutron gamma spectrometer with stilbene scintillator. The pulse shape discrimination technique which would come of the zero cross over method was used to measure the fast neutron and gamma-ray spectra. Thermal neutron fluxes have been measured by using the thermal neutron detection system and the BF-3 detector. The attenuation parameters: precisely, macroscopic effective removal cross-sections ΣR (cm-1), macroscopic cross-sections Σ (cm-1), and total attenuation coefficients μ (cm-1) of fast and thermal neutrons and total gamma-rays respectively were evaluated using the attenuation relations. Also, the MCNP5 code and MERCSF-N program have been used to compute the parameters theoretically. When applicable, measured and calculated results were compared, and it tells us a comprehensive agreement.

The influence of MgO on the radiation protection and mechanical properties of tellurite glasses

Mechanical moduli, such as Young’s modulus (E), Bulks modulus (B), Shear modulus (S), longitudinal modulus (L), Poisson’s ratio (σ) and micro Hardness (H) were theoretically calculated for (100-x)TeO2+x MgO glasses, where x = 10, 20, 30, 40 and 45 mol%, based on the Makishima–Mackenzie model. The estimated results showed that the mechanical moduli and the microhardness of the glasses were improved with the increase of the MgO contents in the TM glasses, while Poisson’s ratio decreased with the increase in MgO content. Moreover, the radiation shielding capacity was evaluated for the studied TM glasses. Thus, the linear attenuation coefficient (LAC), mass attenuation coefficient (MAC), transmission factor (TF) and half-value thickness (Δ0.5) were simulated for gamma photon energies between 0.344 and 1.406 MeV. The simulated results showed that glass TM10 with 10 mol % MgO possess the highest LAC and varied in the range between 0.259 and 0.711 cm−1, while TM45 glass with 45 mol % MgO possess the lowest LAC and vary in the range between 0.223 and 0.587 cm−1 at gamma photon energies between 0.344 and 1.406 MeV. Furthermore, the BXCOM program was applied to calculate the effective atomic number (Zeff), equivalent atomic number (Zeq) and buildup factors (EBF and EABF) of the glasses. The effective removal cross-section for the fast neutrons (ERCSFN, ∑R) was also calculated theoretically. The received data depicts that the lowest ∑R was achieved for TM10 glasses, where ∑R = 0.0193 cm2 g−1, while TM45 possesses the highest ERCSFN where ∑R = 0.0215 cm2 g−1.

Development of SiO2 based doped with LiF, Cr2O3, CoO4 and B2O3 glasses for gamma and fast neutron shielding

Abstract In this study, the fast neutron and gamma-ray absorption capacities of the new glasses have been investigated, which are obtained by doping CoO,CdWO4,Bi2O3, Cr2O3, ZnO, LiF,B2O3 and PbO compounds to SiO2 based glasses. GEANT4 and FLUKA Monte Carlo simulation codes have been used in the planning of the samples. The glasses were produced using a well-known melt-quenching technique. The effective neutron removal cross-sections, mean free paths, half-value layer, and transmission numbers of the fabricated glasses have been calculated through both GEANT4 and FLUKA Monte Carlo simulation codes. Experimental neutron absorbed dose measurements have been carried out. It was found that GS4 glass has the best neutron protection capacity among the produced glasses. In addition to neutron shielding properties, the gamma-ray attenuation capacities, were calculated using newly developed Phy-X/PSD software. The gamma-ray shielding properties of GS1 and GS2 are found to be equivalent to Pb-based glass.

An extensive study on nuclear shielding performance and mass stopping power (MSP)/projected ranges (PR) of some selected granite samples

ABSTRACT In the present investigation, mass attenuation coefficients () and some related nuclear shielding quantities of various granite samples have been computed at several photon energies of 0.356, 0.662, 1.173, 1.274 and 1.333 MeV by using standard simulation geometry of the Monte Carlo N-particle Transport Code System – extended (MCNPX) code. The estimated data have been compared to those of experimental and NIST standard theoretical. The results are observed to be comparable with each other. This study represented that G1 and G9 granites possess the highest and lowest half-layer value (HVL) among the selected granite samples, respectively. The values of effective atomic number (Z eff) of all the granites were found to be almost stationary in the studied energy region. In addition, exposure buildup factor (EBF) values of the granites were computed for the 0.015–15 MeV energy range up to 40 mfp. The largest and lowest values for EBF were obtained for G8 and G2 samples, respectively. The HVLs of nine granite samples were compared with those of concretes, rocks and polyethylene composites at different photon energies. Additionally, mass stopping power (MSP) and projected range (PR) are calculated using the Stopping and Range of Ions in Matter code. The effective removal cross-sections for fast neutrons (ΣR ) have been also assessed to survey the neutron protection effectiveness of the granites. The results represented that G9 has the best neutron attenuation features. It can be concluded from this study that the G9 granite sample possesses superior gamma and neutron shielding performances.

Evaluation of Radiation Shielding Features of Co and Ni-Based Superalloys Using MCNP-5 Code: Potential Use in Nuclear Safety

Due to their excellent heat resistance, superalloys are used predominantly in the manufacturing of engine parts and accessories for aircraft and aerospace equipment. The Monte Carlo simulation (MCNP-5) code was performed to estimate the mean track length of the incident photons inside six different alloys. Then, based on the simulated track length, other important γ-ray shielding parameters were calculated. In this study, the highest mass attenuation coefficient was obtained for alloys encoded MAR-302 and MAR-247 and varied in the range 0.035–72.94 and 0.035–71.98 cm2·g−1, respectively. The lowest mass attenuation coefficient was found for alloys coded Inconel-718 and Nimocast-75 with a range of 0.033–59.25 and 0.32–59.30 cm2·g−1, respectively. Use was made of a recently developed online program Phy-X/PD to calculate the effective atomic number, equivalent atomic number, and the buildup factors for the alloys of interest. The effective removal cross-section for the fast neutron was also calculated for the studied alloys: the highest value was found for the alloys coded with Inconel-718 (∑R = 0.01945 cm2·g−1) and Nimocast-75 (∑R = 0.01940 cm2·g−1), and the lowest value was obtained for alloy coded MAR-302 (∑R = 0.01841 cm2·g−1). Calculated data indicate that MAR-302 and MAR-247 are superior candidates for shielding of gamma-rays, while Inconel-718 and Nimocast-75 MAR-302 are suitable for the shielding of fast neutrons.

PbO–Sb2O3–B2O3–CuO glassy system: Evaluation of optical, gamma and neutron shielding properties

For a PbO–Sb2O3–B2O3–CuO glass system, optical, gamma and neutron shielding properties have been comprehensively examined. The absorbance of the recent glasses positively correlated to the increase in PbO content. Our results show that both the band gap and refractive index remain nearly unchanged with PbO content. Adopting NISTXCOM within an energy photon range of 0.015–15 MeV, the shielding parameters were evaluated for glass samples. The results from MAC outline good compliance with the obtained values using FLUKA codes and Phy-X/PSD. Comparatively higher density (4.81 g/cm3), greater shielding parameters values were all achieved for 35PbO–5Sb2O3–60B2O3-0.5CuO (PSBC5). 31.2, 35.6, 44.9, 55.1, and 66.3 cm2/g values are the mass attenuation coefficient of PS PSBC1, PSBC2, PSBC3, PSBC4, and PSBC5 glasses at 0.015 MeV. This indicating that PSBC5 sample is the superior gamma shield. In addition, the calculated effective removal cross-sections for fast neutrons observe that the PSBC5sample has commensurately greater ΣR (=0.113 cm−1).

Experimental and Monte Carlo simulation study on potential new composite materials to moderate neutron-gamma radiation

In this study, 12 different concentrations of shielding materials were developed and produced. They were covered with high temperature resistant (1500 °C) sodium silicate sealing paste. Epoxy resin was produced by adding different percentages of additive materials such as chromium oxide (Cr2O3), lithium (LiF), and nickel oxide (NiO). The GEANT4 and FLUKA codes of the Monte Carlo simulation toolkit were used to determine the mixing ratios. The total macroscopic cross-sections, effective removal cross-sections, mean free path, half value layer, and transmission neutron number were determined for fast neutron radiation using GEANT4 and FLUKA simulation codes. The mass attenuation coefficient, the effective atomic number and half-value layer (HVL) of the samples were calculated using Phy-X/PSD software. The absorbed dose was measured. In this study, an 241Am–Be neutron source with 74 GBq activity and average neutron energy of approximately 4.5 MeV and a BF3 gas detector were used. Both simulation and experimental measurements were compared with paraffin and conventional concrete. The new composite shielding material absorbed radiation much better than the reference materials. This new radiation shielding composite material can be used in nuclear medicine, transport and storage of radioactive waste, nuclear power plants, and as a shielding material for neutron and gamma radiation.

Radiation Shielding Investigation of Lead Borophosphate Glasses Modified by Tungsten Oxide

Abstract The effect of tungsten trioxide (WO3) on the shielding properties of the lead(II) oxide–boron trioxide–phosphorus pentoxide–WO3 system was investigated. The mass attenuation coefficients (μ/ρ) were evaluated using the XCOM program (National Institute of Standards and Technology, Gaithersburg, MD) in the energy range of 0.015–10 MeV. The μ/ρ values were then used to calculate the mean free path (MFP) and half value layer (HVL). In addition, the macroscopic effective removal cross section for fast neutrons (ΣR) was calculated for the present glass system. It was found that the μ/ρ values increase with increasing WO3 concentration, while HVL and MFP values decrease with increasing WO3 content. The present glass system showed good radiation shielding properties by providing shorter MFP than some glass systems, such as 70TeO2-30V2O5 and 80TeO2-20MoO3 glasses. In addition, it was found that the ΣR values for the present glasses lie within the range of 0.1100–0.1132 cm−1.

Investigating the effect of adding CdO nano particles on neutron shielding efficacy of HDPE

Novel neutron shielding material based on borated high-density polyethylene (HDPE) was prepared in this study. In order to enhance the shielding efficacy of borated HDPE, nano particles of high neutron capture cross section material i.e. of cadmium oxide (CdO) were used as neutron absorber. The CdO nano particles were prepared with composite hydroxide mediated (CHM) method and tested with X-ray diffraction to ensure their structural parameters. The grain sizes of CdO sample obtained from reflection peaks at 2θ = 30.39° and 23.52° were 24.9 nm and 36.1 nm, respectively. The corresponding hkl values and lattice spacing for above-mentioned reflection peaks were found (111) & (110) and 0.29 & 0.38, respectively. These nano scale CdO particles were then mixed in HDPE/B4C composite (containing 10% by wt. of B4C) in predetermining concentrations (i.e. 1%, 2%, and 4% by wt.). The admixture were then pressed in sheets and tested for structural stability and neutron shielding efficacy. The structural stability analysis with FTIR revealed the fact there was no significant alteration in HDPE structure on addition of CdO. The values of effective mass removal cross sections from neutron shielding measurements were 0.1567cm2g-1, 0.1906cm2g-1, and 0.1634cm2g-1 for composites containing 1%, 2%, and 4% CdO, respectively. Furthermore, among all, composite with 2% CdO has been observed of having best shielding properties as it contained the appropriate amount of all the constituents. The higher neutron capture cross section of CdO nano particles was attributed for this significant increase in shielding efficacy of HDPE/B4C/CdO nano composites as compared to pure HDPE and HDPE/B4C composites.

Structural and radiation shielding properties of BaTiO3 ceramic with different concentrations of Bismuth and Ytterbium

This study aims to explore the effect of Bi2O3 and Yb2O3 on the structural and radiation shielding properties of BaTiO3 ceramic samples fabricated by solid-state reaction and high-energy ball milling techniques. X-ray diffraction (XRD) spectra revealed that the current ceramic samples have a tetragonal phase. Fourier Transform Infrared (FTIR) results showed a peak at 1435 cm−1 for all ceramics that attributed to Ba–Ti–O. Various physical properties like density, packing density, molar volume, and Poisson's ratio were calculated for the prepared ceramic samples. The highest concentration of Bi2O3 (C3) exhibited the highest density values (5.562 g/cm3), packing density (0.630), and Poisson's ratio (0.279), whereas the high concentration of Yb2O3 (C6) had a density of 4.731 g/cm3, a packing density of 0.529, and a Poisson's ratio of 0.237. In addition, different radiation shielding properties were determined by using the Phy-X program for all ceramic samples. The mass attenuation coefficient (μ/ρ) values for C3 and C6 at 1 MeV were 0.065 and 0.062 cm2/g, respectively. Based on the calculated (μ/ρ), the C3 sample revealed a promising half-value layer (HVL), mean free path (MFP), and effective atomic number (Zeff). Regarding the neutron shielding properties, the effective removal cross-section (ΣR) values for C1, C2, C3, C4, C5 and C6 were 0.101, 0.098, 0.095, 0.089, 0.88 and 0.085 cm−1, respectively. In this regard, C1 showed the highest ΣR compared with the other prepared samples. In conclusion, C1 and C3 samples have superior neutron and photon shielding features compared with the other prepared ceramics. In light of these obtained results, the fabricated ceramics can be concluded to be very helpful for photon and neutron shielding applications.

Nuclear shielding properties of B2O3–Pb3O4–ZnO glasses: Multiple impacts of Er2O3 additive

This study aimed to examine nuclear radiation shielding properties of a novel glass system, which has nominal compositions (40-x) B2O3 +40Pb3O4 +20ZnO + xEr2O3 with several substitution proportions of Er2O3 (x = 0, 1, 2, 3, 4 and 5). The conventional solid-state method was utilized to fabricate the Er2O3 doped glasses. To obtain the amorphous nature of manufactured glass samples, X-ray diffraction (XRD) patterns were investigated. Moreover, mass attenuation coefficients (μ/ρ) of the prepared glasses were adjusted by employing MCNPX simulation code. Other important gamma-ray attenuation parameters namely Half Value Layer (HVL), Mean Free Path (MFP), Effective and Equivalent Atomic Numbers (Zeff and Zeq) and Exposure Buildup Factor (EBF) of the glasses were also extensively addressed by obtaining mass attenuation coefficients (μ/ρ). The outcomes indicated that the Er2O3 additive enhances the photon attenuation qualification of the glasses. Moreover, the success of glasses in reducing fast neutrons was judged by calculating effective removal cross sections (ΣR). The results showed that increasing Er2O3 additive upgraded the prepared glasses in the attenuation of neutrons as in gamma rays. In addition, the protection ability of the fabricated glasses against alpha and proton charged particles was also evaluated. Charged alpha and proton particles in BPZE5 glass have the shortest projected ranges which are 50.38 μm and 436.36 μm respectively. It was seen that the increase of Er2O3 additive upgraded the prepared glasses in terms of attenuation of all of the selected radiation types. It can be deduced that these results might help for estimating the effect of Er2O3 on different glass systems in the future.

Investigation of photon, neutron and proton shielding features of H3BO3–ZnO–Na2O–BaO glass system

The current study aims to explore the shielding properties of multi-component borate-based glass series. Seven glass-samples with composition of (80-y)H3BO3–10ZnO–10Na2O–yBaO where (y = 0, 5, 10, 15, 20, 25 and 30 mol.%) were synthesized by melt-quench method. Various shielding features for photons, neutrons, and protons were determined for all prepared samples. XCOM, Phy-X program, and SRIM code were performed to determine and explain several shielding properties such as equivalent atomic number, exposure build-up factor, specific gamma-ray constants, effective removal cross-section (ΣR), neutron scattering and absorption, Mass Stopping Power (MSP) and projected range. The energy ranges for photons and protons were 0.015–15 MeV and 0.01–10 MeV, respectively. The mass attenuation coefficient (μ/ρ) was also determined experimentally by utilizing two radioactive sources (166Ho and 137Cs). Consistent results were obtained between experimental and XCOM values in determining μ/ρ of the new glasses. The addition of BaO to the glass matrix led to enhance the μ/ρ and specific gamma-ray constants of glasses. Whereas the remarkable reductions in ΣR, MSP, and projected range values were reported with increasing BaO concentrations. The acquired results nominate the use of these glasses in different radiation shielding purposes.

Assessment of gamma-ray attenuation features for La+3 co-doped zinc borotellurite glasses

In this study, gamma ray and neutrons shielding features were evaluated for La+3 co-doped zinc borotellurite glasses with chemical formula {[(TeO2)0.70(B2O3)0.30]0.7(ZnO)0.30}1-x(La2O3)x: 0.01≤ x ≤ 0.05 mol%. Firstly, the mass and linear attenuation coefficients (MAC, LAC) for glasses were calculated via Photon Shielding and Dosimetry (PSD) software (Phy-X/PSD) in the energy range of 0.015–15 MeV. Depending on MAC and LAC, effective shielding parameters: Half value layer and mean free path (HVL, MFP), effective electron density, atomic number, and conductivity (Neff, Zeff, Ceff), exposure and energy absorption buildup factors (EBF, EABF) have been evaluated. In addition, effective fast neutron removal cross section (FNRCS) for the studied glasses have been calculated. The obtained data were compared with those of some commercial glasses and concrete. Results reveal that HVL for all samples can be arranged in the order TBZL1>TBZL2>TBZL3>TBZL4>TBZL5. The HVL, TVL and MFP values of TBZL5 glass are 2.988 cm, 9.927 cm, and 4.311 cm, respectively. TBZL5 glass has the highest Ceff values and TBZL1 glass has the lowest Ceff values. EBF values change range from 1 to 4 at 1 MFP and from 1-1010 at 40 MFP value. Therefore, it can be among the alternative materials that can be preferred in radiation application areas.

Scanning electron microscopy (SEM), energy-dispersive X-ray (EDX) spectroscopy and nuclear radiation shielding properties of [α-Fe3+O(OH)]-doped lithium borate glasses

Goethite [α-Fe3+O(OH)]-doped lithium borate glasses in the chemical form of [(100 − x) Li2B4O7 mineral x = 10, 20, 30, 40 wt%] were synthesized and investigated in terms of their ability to protect the human being and environment from ionizing radiation. Elemental analysis of four different goethite [α-Fe3+O(OH)]-doped lithium borate glass samples called GOG10, GOG20, GOG30 and GOG40 prepared in different contribution rates was tested using energy-dispersive X-ray technique. In addition, the surface morphology of the manufactured glass samples was determined by scanning electron microscopy. The glass samples subject to the study were evaluated in terms of radiation shielding properties by calculating the shielding parameters such as mass attenuation coefficient (μm), half value layer, mean free path, effective atomic number (Zeff), exposure and energy absorption build-up factors, effective removal cross section (ΣR) for fast neutron and RDD (for gamma and neutron) with the help of the XCOM and MCNPX simulation code. Mass stopping power and projected ranges values of the studied glass samples were calculated in order to consider the shielding performance against proton and alpha particles. The results showed that increase in goethite additive increases the gamma protection capacity of glasses. As a result, the glass sample with the highest goethite contribution encoded with GOG40 has better shielding efficiency in terms of the gamma and neutron radiation shielding.

Newly developed Zinc-Tellurite glass system: An experimental investigation on impact of Ta2O5 on nuclear radiation shielding ability

In this paper, the nuclear shielding properties of newly fabricated Zinc-Tellurite glasses doped with Ta2O5 with a composition (25ZnO.75TeO2)100-x.(Ta2O5)x (x=0, 1, 2, 3 mol%) were studied. Ta2O5 doped Zinc-Tellurite glasses prepared by the traditional melt quenching method. The mass attenuation coefficients were achieved for prepared glasses at 0.081-0.383 MeV photon energies employing experimental transmission measurements, which were performed with 3 Ci Ba-133 point source and Ultra Ge detector. The experimental results were compared with simulated MCNPX and theoretical WinXCOM data. Moreover, another important nuclear shielding parameters such as mass attenuation coefficient, half value layer, effective atomic number, effective electron density, equivalent atomic numbers, alpha and proton projected ranges, mass stopping power in the protection of gamma, neutron, alpha and proton radiation were calculated. In addition, buildup factors of investigated glasses were determined by using the G-P fitting method. Moreover, effective removal cross section ΣR and transition factors were determined. According to the results, the addition of Ta2O5 to the glass samples has a positive effect on the performance of nuclear shielding effects of the glass samples. It can be concluded that ZTT3 glass sample with the highest Ta2O5 contribution has excellent nuclear radiation shielding ability among the other fabricated glass samples.

Comparison of gamma and neutron shielding competences of Fe–Cu- and brass-added Portland cement pastes: an experimental and Monte Carlo study

In this study, Portland cement paste samples containing (Fe–Cu)x and brassx (where in x = 0, 5, 10, 15, 20%) were manufactured and some physical and nuclear shielding properties were determined. Experimental measurements for determining the mass attenuation coefficients (μ/ρ) of the cement samples were accomplished by utilizing 133Ba radioactive source with HPGe detector. In addition, gamma transmission factor and μ/ρ values were simulated with MCNPX codes and the experimental results were checked with MCNPX and theoretical WinXCOM results. The largest μ/ρ values were achieved by 20% Fe–Cu- and brass-added samples varying between 0.344–0.098 and 0.363–0.099 cm2/g for 0.081–0.383 photon energy ranges, respectively. To extensively evaluate the photon shielding competences of the cement samples, other vital parameters such as HVL, MFP, Zeff and Nel derived from μ/ρ values were also calculated. It is seen that the highest Zeffs were obtained for 20% brass- and Fe–Cu-added samples, whereas for MFP and HVL and Nel values, the least values were obtained. Additionally, to estimate the neutron shielding effectiveness of the cement samples, effective removal cross section (∑R) values were obtained. The ∑R values varied between 0.1123–0.1349 and 0.1065–0.1361 cm−1 for Fe–Cu- and brass-doped samples, respectively. The data obtained from the current study showed that adding Fe–Cu and brass improves the nuclear shielding properties of the produced cement samples.