Effective Removal Cross-section Research Articles

Effect of using heavy aggregates on the high performance concrete used in nuclear facilities

For nuclear power plants construction, each country has to make use of its own available raw materials. Therefore, through the current research we had to search for the relevant local materials suitable for usage as concrete components and satisfy the needed requirements. The target of this research is to investigate the effect of some type’s heavy aggregates on the physico-mechanical properties and attenuation coefficient of the high performance heavy concrete used as radiation shielding of nuclear facilities. For this purpose radiation shielding properties of high performance heavy concrete mixture prepared using different types of heavy aggregates such as magnetite in addition to industrial waste of lead slag as a coarse aggregate and fine aggregate of local sand or replacing sand by fine magnetite or lead slag, were investigated comparatively with hardened dolomite concrete. The physico-mechanical properties of type aggregates and hardened concretes were examined. Gamma transmission parameters for the studied concrete for Cs137 and Co60 as gamma source were measured, also the effective removal cross-section (ΣR) of fast neutrons spectra for a wide range of neutron energies of [2–12 MeV] with thicknesses 4, 6, 8, 10 and 12 cm of heavy concrete mixes for Plutonium-Beryllium as neutrons source. While the thermal neutrons tests were performed using californium radioisotope. An innovative material that the current research has introduced to be used as a heavy coarse aggregate in HPC is “lead slag”, lead slag has been converted from a subsidiary industrial and environment polluting waste to a very valuable material as one of the ingredients of heavy concrete. It was found that, the magnetite and lead slag aggregates improved the physico-mechanical properties of heavy concrete compared to dolomite concrete. Replacing fine sand by the fine aggregates of magnetite or lead slag enhances the efficiency of attenuation by the concrete mixes against these radiations and the compressive strength values increase by 12 and 27% respectively, of heavy concrete after 28 days. The concrete mix containing lead slag aggregate has higher mechanical properties and enhances the attenuation properties for fast neutrons, gamma rays and thermal neutrons than the other concrete mixes.

Fabrication of novel neutron shielding materials: Polypropylene composites containing colemanite, tincal and ulexite

Exposure to neutron radiation can cause serious health problems, especially in nuclear power plants and their surroundings, high-flying aircrafts, and healthcare facilities. Conventionally, many neutron-shielding materials are used for protection, but their mechanical and thermal properties limit the use of these materials. Low density, low cost, and flexible neutron shields are becoming increasingly important for these uses, in particular where the weight factor, such as in the aerospace sector, is highly important. For this reason, polypropylene (PP) based composite materials were prepared and their neutron shielding features, such as effective removal cross-sections, half value layer, and mean free path were determined using GEANT4 code. The absorbed dose-measuring experiments were carried out using a241Am-9Be neutron source. Boron minerals such as colemanite, tincal, and ulexite were used in the production of the materials to reduce costs and increase neutron absorption capacity. All the results were compared with the shielding abilities of paraffin and conventional concrete. It was determined that these composites absorb neutron radiation much better than the referenced materials. Colemanite was characterized by X-ray diffraction (XRD) while PP was characterized by Fourier transform infrared (FTIR) spectroscopy and differential scanning calorimetry (DSC). These composites prove to be valuable shielding materials for protection against neutron radiation in nuclear medicine, storage or transportation of radioactive waste, and other nuclear applications.

The Impact of PbF2-Based Glasses on Radiation Shielding and Mechanical Concepts: An Extensive Theoretical and Monte Carlo Simulation Study

This work aimed to investigate the impact of Lead-fluoride based glasses via theoretical and simulation techniques on mechanical and radiation shielding parameters. Accordingly, Bi2O3 was blended with TeO2–B2O3–PbF2 glasses by using melt-quenching method. Using Fluka Monte Carlo code, the radiation shielding properties have measured. Moreover, Comparatively higher density PbF80 = 6.163 g/cm3 with 80 mol % Bi2O3, greater µ, µm and Zeff and lower T1/2, λ, tenth value layer values achieved for TeO2–B2O3– PbF2/Bi2O3 glass pointed it out as the best shield of gamma. Besides, the computed effective removal cross-sections against fast neutrons (ΣR) observed that the PbF80 sample has commensurately greater with a value 5.2954 (cm−1). The results observed that the variation PbF2/Bi2O3 improves the gamma protection ability of Lead-fluoride based glasses. The incremental of Bi2O3 at the expense of PbF2 increases the density and the packing density of the glasses and hence increase the longitudinal modulus-L, shear modulus-S, bulk modulus-K, and Young's modulus-Y from 15.89 to 25.9 GPa, from 8.49 to 12.09 -GPa, from 4.58 to 9.77 GPa, and from 15.74 to 25.69 GPa, respectively. The increase of the elastic moduli can be attrinuted to the glass former role of Bi2O3. The results indicate that the highest PbF2/Bi2O3 ratio encoded PbF00 has the best shielding and mechanical competence with measurable physical properties.

Novel HMO-Glasses with Sb2O3 and TeO2 for Nuclear Radiation Shielding Purposes: A Comparative Analysis with Traditional and Novel Shields.

The radiation shielding characteristics of samples from two TeO2 and Sb2O3-based basic glass groups were investigated in this research. TeO2 and Sb2O3-based glasses were determined in the research as six samples with a composition of 10WO3-(x)MoO3-(90 − x)(TeO2/Sb2O3) (x = 10, 20, 30). A general purpose MCNPX Monte Carlo code and Phy-X/PSD platform were used to estimate the radiation shielding characteristics. Accordingly, the linear and mass attenuation coefficients, half value layer, mean free path, variation of the effective atomic number with photon energy, exposure and built-up energy factors, and effective removal cross-section values were determined. It was determined that the results that were produced using the two different techniques were consistent. Based on the collected data, the most remarkable findings were found to be associated with the sample classified as T80 (10WO3 + 10MoO3 + 80TeO2). The current study showed that material density was as equally important as composition in modifying radiation shielding characteristics. With the T80 sample with the greatest density (5.61 g/cm3) achieving the best results. Additionally, the acquired findings were compared to the radiation shielding characteristics of various glass and concrete materials. Increasing the quantity of MoO3 additive, a known heavy metal oxide, in these TeO2 and Sb2O3-based glasses may have a detrimental impact on the change in radiation shielding characteristics.

Ionizing radiation shielding features for titanium borosilicate glass modified with different concentrations of barium oxide

Photon, proton, neutron and alpha shielding features of 10TiO2–10SiO2-(80-x)B2O3-xBaO were studied. This study exhibits several photon shielding parameters such as mass attenuation coefficient (MAC), half-value layer (HVL), effective atomic number (Zeff), effective electron density (Neff), gamma dose rate (R/h), specific gamma-ray constant (Γ), and the specific absorbed fraction of the energy (SAFE). The effective removal cross-section of the fast neutron (ΣR), scattering length of the constituent elements, mass stopping power (MSP), and projected range variation parameters were used to explore the particle shielding features. Moreover, the MAC was determined by using Geant4 simulation and compared with the results of the Phy-X program. The percentage error (Δ) was less than 2% for all results, which revealed the agreement between Geant4 simulation and Phy-X program. Although, the addendum of barium oxide (BaO) led to improving the performance of MAC, HVL, Zeff, Neff, and Γ. The ΣR, MSP and projected range variation showed a progressive reduction with incorporation BaO. Lastly, the BaO and SiO2 played an amazing role in enhancement the durability for the glass sample, and the S4 sample exhibits the highest durability compared with other samples. From the obtained results, it can be concluded that the present glass samples can be used in the radiation-shielding field.

Evaluation of the physical, structural, thermal, and advanced radiation absorption characteristics of SiO2–Na2O–K2O–P2O5–CaO bioactive glasses

We report herein an evaluation of the physical, structural, thermal, advanced photon energy absorption, and fast neutron removal characteristics of a series of bioactive glasses having the compositions 20SiO2-7.5Na2O-7.5K2O-(65–x)P2O5-xCaO, where x = 2.5, 5.0, 7.5, and 10.0 mol.%. X-ray diffraction (XRD) patterns confirmed the amorphous nature of these phosphosilicate glass systems fabricated by the melt-quench technique. To confirm the in vitro bioactivity of the prepared glasses, powdered samples of each composition were compressed into pellets and then soaked in simulated body fluid (SBF) at 37 ± 0.5 °C for 28 days. Soaked samples were investigated by means of XRD, scanning electron microscopy (SEM), and energy-dispersive X-ray spectroscopy (EDS/EDX). Thermo-gravimetric and differential thermal analysis (TG-DTA) was applied to determine the glass transition temperature (Tg) and crystallization temperature (Tc) for each glass sample. Fourier-transform infrared (FTIR) measurements were performed in the range 4000–400 cm−1, and showed the bending and stretching motions of different groups and the presence of CO2 and OH− groups in the glasses. Radiation absorption properties, such as mass energy absorption coefficient (μen/ρ), kinetic energy released per unit mass (KERMA) relative to air (ҠR), and mean free path values have been obtained for the bioactive glasses and compared with those of registered 45S5 Bioglass®. The simultaneous variation of mass energy absorption and mass attenuation values with photon energy has been explored to visualize the difference between absorption and attenuation. Linear attenuation coefficients for the present glass samples have been measured at photon energies of 511, 662, 1173, and 1332 keV in narrow-beam transmission geometry, and were in good agreement with theoretical results (error < 2 %). Further, detailed comparative analysis of energy absorption build-up factor (EABF) values for the glass samples has been conducted using two software tools, namely Photon Shielding and Dosimetry (PSD) and EXABCal, as well as on the basis of equivalent atomic number (Zeq) and effective atomic number (Zeff). We also attempted to calculate effective removal cross-sections (∑R) in order to evaluate the neutron absorption performances of these glass samples. The present detailed analysis shows that glass sample S4 with 10 mol.% CaO exhibits the most promising radiation absorption properties, resembling those of 45S5 Bioglass®.

Study on recycled Er-incorporated waste CRT glasses for photon and neutron shielding

The present study is extremely associated with the evaluation of cathode ray tube (CRT) waste which is a growing hazardous waste management and disposal issue. The Er-added recycled glasses derived from CRT have been successfully fabricated as a component for Na2O–Al2O3–SiO2–K2O–CaO–SrO–BaO–PbO–Fe2O3–TiO2–Er2O3 glass systems. The waste cathode ray tube panel glass (WPG) series were obtained by inserting the Er2O3 in substitution for CRT in the amounts of 0, 1, 3 and 5 mol%. The synthesized WPG series were then subjected to numerous characterization analysis. One can firmly report that the insertion ratio from 0 to 5 mol% in Er2O3 increased the glass density from 2.9216 to 2.9763 g/cm3. Further, there exists no crystalline formation with the addition of Er2O3, instead, an amorphous nature in all WPG series emerges. Microstructural images captured by the SEM technique also confirm the non-crystallinity of the fabricated samples. Energy dispersive spectroscopy (EDS) in SEM, besides, paved the way for revealing the elemental mapping of the WPG series. From the perspective of radiation shielding competencies, nuclear shielding features have been evaluated in terms of μm and related attenuation parameters at several photon energies of 276.4–1332.5 keV by utilizing narrow beam transmission methods. The experimental data have been also confirmed to those of theoretical results by WinXCom software. The HVL results revealed superior photon radiation shielding performances for recycled WPG glasses in comparison to various concretes. In addition, exposure buildup factor (EBF) values estimated utilizing the G-P fitting approach would be useful to design or develop the synthesized glass systems for shielding applications. Furthermore, the macroscopic effective removal cross-sections for fast neutron (ΣR) have been estimated. In conclusion, the findings clearly demonstrated that Er-incorporated CRT glasses can effectively be implemented in radiation shielding applications instead of Pb having toxic effects.

Monte Carlo simulation on shielding properties of neutron-gamma from 252Cf source for Alumino-Boro-Silicate glasses

Neutron-gamma attenuation characteristics of the Alumino-Boro-Silicate (ABS) glasses were estimated based on MCNPX simulation and theoretical calculations. The watt fission distribution was used to extract the neutron spectrum and the obtained results were reported in the presence and absence of the glass sample. The other attenuation parameters such as effective removal cross section (ΣR), mass removal cross section (∑Rρ), Half Value Thickness (HVT), Tenth Value Thickness (TVT), and also mean free path (mfp) were also demonstrated. The received data depicts that glass coded ABS-0 is an appropriate material in terms of neutron attenuation purposes in comparison to the other selected shielding materials. Moreover, it can be figured out that the increment of the Bi2O3 rate decreases the neutron shielding ability of the Alumino-Boro-Silicate glass and has a negative effect on the fast neutron shielding proficiency. Besides, using Doppler Effect gamma spectrum for spontaneous fission 252Cf source were extracted in shielding materials, and data represented graphically. Linear attenuation coefficient(μ), HVT, mfp for a wide energy range of incident gamma rays were also estimated. The simulated results showed that the ABS-4 glass sample possesses the highest μ value (better gamma shielding efficiency) which varies in the range between 0.381763 and 0.954323 cm−1 while the lowest μ value is obtained for ABS-0 that varies in the range between 0.339447 and 0.511594 cm−1.

Characterization of gamma-ray and neutron radiation absorption properties of synthesized quinoline derivatives and their genotoxic potential

In this study, six different types of quinoline derivatives were synthesized, which can be used as an anti-radiation drug for active matter and its effects against radiation was determined. The effective removal cross-sections, mean free path, halfvalue layer and neutron transmission number were calculated for fast neutron radiation using the Monte Carlo simulation GEANT4 and FLUKA codes. Gamma radiation shielding parameters such as mass attenuation coefficient (MAC), effective atom number (Zeff), mean free path (MFP), exposure buildup factor (EBF) and halfvalue layer (HVL) were theoretically calculated using WinXCom and Phy-X/PSD software. Neutron equivalent dose measurements were performed using a 241Am–Be fast neutron source and a BF3 gas detector with 4.5 MeV of energy and 74 GBq activity. Both simulation and experimental measurements were compared with paraffin and conventional concrete. It was found that the quinoline derivative shielding material absorbed radiation much better than these reference materials. Additionally, the genotoxic potentials of the derivatives were assessed by using the yeast DEL assay and the results revealed that the derivatives can be considered genotoxically safe at the tested concentrations (up to 5 mM). It has been suggested that this new radiation shielding derivatives material can be used as active ingredients for a drug to be developed against both neutron and gamma radiation.

Repercussions of yttrium oxides on radiation shielding capacity of sodium-silicate glass system: experimental and Monte Carlo simulation study

Five glass samples with composition 37.5 Na2O + (61.1 − x) SiO2 + xY2O3, x = 1, 2, 5, 5.7 and 6 mol% were synthesized using the melting methods at temperatures varying between 1400 and 1500 °C, depending on the Y2O3 insertion ratio. The synthesized glasses density was measured using the Archimedes method. Moreover, the linear attenuation coefficient was measured experimentally using the gamma-ray transmission method with NaI (Tl) crystal and radio-isotopes C-137 and Co-60. The mass attenuation coefficient was calculated based on the experimental linear attenuation coefficient and confirmed using the Monte Carlo simulation MCNP-5 code and XCOM program. The gamma-ray shielding capacity was enhanced with the replacement of SiO2 by Y2O3. The LAC at energy 0.662 MeV varied in the range between 0.201 and 0.232 cm−1, increasing the Y2O3 insertion ratio between 1 and 6 mol%, respectively. The fast neutron effective removal cross section ∑R (cm2/g) was also evaluated theoretically for the fabricated glass samples. The computed ∑R values decreased linearly with the Y2O3 content.

Novel concrete compositions for γ-rays and neutron shielding using WC and B4C

Concrete provides a superior shielding effect against γ-rays and neutron by reducing the radiation exposure to biological entities and their surroundings. A suitable amount of additives in the concrete mixture can enhance the shielding effect. With this perception, the mixing proportion of concrete has been prepared with different amounts of WC and B4C additives. The shielding effect of prepared concretes with WC and B4C additives have been investigated using experimental data (for 60Co and 252Cf sources), and theoretical prediction using codes MCNP6, XCOM, Auto-Zeff, and NXcom. Different shielding parameters such as; mass attenuation coefficient (μm), effective atomic number (Zeff), effective electron density (Neff), half-value layer (HVL), tenth-value layer (TVL), mean free path (MFP), and effective removal cross-section (ΣR) were calculated in the present work. The obtained experimental and MCNP6 results for γ-rays (μm) and neutron (ΣR) are in decent agreement with the difference of less than 6.5% and 8%, respectively. The outcomes of the present work show that the modified compositions have more advantages as compared to the pristine concrete. These also revealed that shielding parameters strongly depend on the atomic composition and density of additives of the prepared concrete. Hence, it has been demonstrated that the addition of an appropriate amount of WC and B4C to concrete may be an alternative option that can be used in several radiation protection applications.

Structural, optical and nuclear radiation shielding properties of strontium barium borate glasses doped with dysprosium and niobium

Five new strontium barium borate (BNBD) glasses doped with dysprosium ion and different concentrations of niobium pentoxide were synthesized using the standard melt-quenching method. The physical, structural, optical and gamma radiation shielding properties of these glasses were investigated. Density, average molecular weight, refractive index, molar volume, optical dielectric constant, boron-boron separation, metallization criterion, oxygen packing density, Poisson ratio, optical basicity, optical electronegativity, and two-photon absorption coefficients of the synthesized glasses were determined. By the addition of Nb2O5 content, boron-boron distance and oxygen packing density values increased, while molar volume of oxygen decreased due to the formation of bridging oxygen. The two-photon absorption could be constrained by replacing BaCO3 by niobium pentoxide content which further influences the bandgap. The ionic nature of the titled glasses is discussed using the bonding parameter, optical basicity, ionic and covalent characteristic parameter values. Moreover, the shielding ability of dysprosium ions doped niobium borate glasses against photons, fast neutrons and electrons has been extensively evaluated. For this purpose, the mass attenuation coefficient (µ/ρ, cm2/g) of the glasses and several photon protection parameters, derived from µ/ρ were obtained for 0.015–15 meV. The maximum µ/ρ values were achieved for BNBD0 glass, varying between 0.033 and 35.430 cm2/g. The lowest buildup factor values were found for BNBD0 glass. Furthermore, effective removal cross section values for fast neutrons increased steadily between 0.125 and 0.130 cm−1 due to the increase in the density of the glasses with the enhancing of Nb2O5 concentration. It was noticed that the range of high energy electrons was shorter on the BNBD0 glass. It was concluded that BNBD0 glass with high BaCO3 concentration can be considered as an alternative material in nuclear radiation shielding applications.

Development of new heavy concretes containing chrome-ore for nuclear radiation shielding applications

In this study, four newly developed chromium ore based new heavy concretes containing different types of minerals (chromium ore, hematite (Fe2O3), titanium oxide (TiO2), limonite (FeO(OH)nH2O), and siderite (FeCO3) and compounds (galena (PbS), chromium oxide (Cr2O3) and manganese oxide (MnO2) were investigated for neutron and gamma shielding effectiveness. It was observed that the developed concretes have strength up to 30 MPa (4351.131 psi) using mechanical stress tests. In addition, temperature resistance tests were carried out and the results were found to be up to desired level. GEANT4 simulation was employed to determine the effective removal cross-section ΣR (cm−1), transmission number mean free path (MFP), and half value layer (HVL) of the concretes. The equivalent dose rate of fast neutron was calculated using Am–Be source and BF3 proportional counter. Gamma-ray shielding properties mass attenuation coefficients, half-value layer, mean free path, and effective atomic number of the concretes were assessed in the energy range between 0.015 and 15 MeV. The obtained results were compared with paraffin, ordinary concretes, and some heavy concretes. Both theoretical and experimental calculations have shown that all new heavy concretes are capable of absorbing high rates of gamma and neutron radiation than reference samples. In particular, the D3 sample has the maximum neutron attenuation dose value, which contains 55% chromium ore. D2 is superior gamma shielding concrete and this contains 60% chromium ore, 10% hematite, 5% TiO2 and 5% PbS. The new developed concretes are found to be alternative shielding material for the nuclear energy applications.

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.