Fast Neutron Cross Sections Research Articles

Production and evaluation of the polyester composites containing pyrite and niobium diboride for radiation protection

The presented study focused on investigating the radiation shielding properties of polyester/pyrite/niobium diboride (FeSNbB) composites. Various gamma-ray shielding parameters were evaluated with help of a HPGe detector system. Theoretical and simulation results are further evaluated by WinXCOM, GEANT4 and FLUKA codes, and the obtained results are compared with experimental values in the energy region 59.5 keV–1332.5 keV. The prepared samples are additionally evaluated using kerma relative to air values and Exposure build-up factors. Then, for the evaluation of neutron radiation shielding, effective removal cross sections for fast neutrons were determined for FeSNbB composites. Electron attenuation efficiency (AE%) for the produced composites was assessed across the energy spectrum of 4 MeV–18 MeV, utilizing experimental data from a Linac machine and theoretical predictions from Eclipse TPS. The results were found to be compatible with each other. FeSNbB50 has the best shielding properties among the studied composites.

Effect of low concentrations of WO3 on synthesis, structural, physical, optical, neutron attenuation, and charged particle absorption properties of B2O3+Na2O+BaO glassy composites

This study presents the physical, optical, and radiation shielding attributes of 75B2O3+5Na2O+(20-x)BaO+xWO3 glasses, with the aim of producing effective transparent and light shields. The well-known melting-and-quenching technique was used to prepare the transparent glasses for x = 0; 03; 10 mol%). The glasses were characterized by determining their density, molar volume, optical transmittance, bandgap, refractive index, dielectric constant, reflection loss, molar refraction charged radiation stopping powers, projected range, fast neutron removal cross section and thermal neutron total cross-section. All the estimated parameters showed variations with glass stoichiometry. The addition of WO3 triggered the BO4 → BO3 unit transformation which prompted the variations in the physical and optical parameters of the glasses. The optical transparency in the visible region and density of the glasses decreased with WO3 concentration while the molar volume increased. For electrons, the stopping powers were in the range, 1.436–11.76 cm2/g, 1.433–11.66 cm2/g, and 1.426–11.44 cm2/g, for BNB-W00, BNB-W03, and BNB-W10, respectively. For protons, alpha particles, and carbon ions, the mass stopping powers increased with WO3 content. Based on projected range data, particles with greater energy per nucleon penetrated the glasses deeper and the WO3- rich glasses had higher transmissivity for charged radiation. Compared to some existing neutron moderators, the investigated glasses displayed higher fast neutron cross-sections. The studied glasses can function well as barriers to moderate fast neutrons, attenuate thermal neutrons, and absorbed charged radiation.

Lead fluoride as a multifunctional modifier in bismuth gadolinium borate glasses: Optical, structural, and radiation attenuation characterization

This study systematically evaluates the role of PbF2 in altering the optical and radiation shielding properties of bismuth gadolinium borate glasses with compositions 30Bi2O3–Gd2O3-xPbF2-(70-x)B2O3, for x ranging from 0 to 40 mol. %. The introduction of PbF2 significantly modified the absorbance profile, exhibiting distinct absorption peaks at 3.45, 4.15, and 4.69 eV, shifting from a flat curve in the undoped sample to well-defined peaks with increasing PbF2 content. Crucially, a reduction in both the direct and indirect optical band gaps was observed as the PbF2 content increased, indicating changes in the electronic structure of the glasses. Photon and neutron shielding capabilities were analyzed, revealing that the linear attenuation coefficient (μ), half-value layer (T0.5), and effective atomic number (Zeff) were enhanced with PbF2 additions. Notably, the BGPB5 sample contained 40 mol. % PbF2 outperformed other compositions, offering the highest attenuation efficiency, which surpassed some conventional glasses and concretes. Simultaneously, an inversely proportional relationship was found between PbF2 concentration and the effective removal cross-section for fast neutrons (ΣR), demonstrating the BGPB1 sample's superiority as a neutron shield. The incorporation of PbF2 in the glass matrix not only improved gamma radiation shielding but also tuned the optical properties, marking these glasses as potential candidates for multifunctional radiation shielding applications.

Effect of ZrO2 on Radiation Permeability Properties of Polypropylene

The study investigates the radiation permeability properties including mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), tenth value layer (TVL), half value layer (HVL), fast neutron cross section (FNRC), and mean free path (MFP) of polypropylene (PP) polymer, as well as the produced polymer matrix composites (PP+5% ZrO2, PP+10% ZrO2, PP+15% ZrO2). The studied materials were examined by considering their effect on radiation permeability against gamma and neutron radiation. Additionally, powder size, Archimedes principle (density), XRD, DSC, ATR, and DTA-TG analyses were performed. According to the radiation permeability results of the studied four materials, PP + 15% ZrO2 was found to have the highest LAC values, while PP was found to have the lowest LAC values. The FNRC values of the PP, PP+5% ZrO2, PP+10% ZrO2, and PP+15% ZrO2 materials were found to be 10.038 cm-1, 12.651 cm-1, 15.002 cm-1, and 17.091 cm-1, respectively. The most suitable material for gamma and neutron shielding was found to be 15% ZrO2 reinforced material.

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

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

Fast neutrons and gamma rays shielding properties for some nylon composite

In this study, The removal parameters of macroscopic fast neutron cross-section, linear gamma ray attenuation, mean free path, and half-value layer (half-value thickness) were calculated for a polymer composite containing nylon 6 as a base material and starch powder, silica oxide, and calcium carbonate as reinforcement materials for different energies (0.05, 0.1, 0.3, 0.6, 0.8, 1, 3, 7, 10, 15) and for various concentrations (5, 10, 15, 20, 25). We used the x-com program to calculate gamma ray attenuation values. The results showed that the macroscopic cross-section of fast neutron attenuation and the linear attenuation coefficient for gamma rays increase with increasing reinforcement materials concentration, while the values of half-value layer decrease with increasing concentrations. When the energy increases, the half-value layer and the mean free path also increase, however, the values of the linear attenuation coefficient decrease.

Enhancing Gamma-Neutron Shielding Effectiveness of Polyvinylidene Fluoride for Potent Applications in Nuclear Industries: A Study on the Impact of Tungsten Carbide, Trioxide, and Disulfide Using EpiXS, Phy-X/PSD, and MCNP5 Code

Background: Radiation protection is crucial in various fields due to the harmful effects of radiation. Shielding is used to reduce radiation exposure, but gamma radiation poses challenges due to its high energy and penetration capabilities.Materials and Methods: This work investigates the radiation shielding properties of polyvinylidene fluoride (PVDF) samples containing different weight fraction of tungsten carbide (WC), tungsten trioxide (WO<sub>3</sub>), and tungsten disulfide (WS<sub>2</sub>). Parameters such as the mass attenuation coefficient (MAC), half-value layer (HVL), mean free path (MFP), effective atomic number (Z<sub>eff</sub>), and macroscopic effective removal cross-section for fast neutrons (Σ<sub>R</sub>) were calculated using the Phy-X/PSD software. EpiXS simulations were conducted for MAC validation.Results and Discussion: Increasing the weight fraction of the additives resulted in higher MAC values, indicating improved radiation shielding. PVDF–xWC showed the highest percentage increase in MAC values. MFP results indicated that PVDF–0.20WC has the lowest values, suggesting superior shielding properties compared to PVDF–0.20WO<sub>3</sub> and PVDF–0.20WS<sub>2</sub>. PVDF–0.20WC also exhibited the highest Z<sub>eff</sub> values, while PVDF–0.20WS2 showed a slightly higher increase in Z<sub>eff</sub> at energies of 0.662 and 1.333 MeV. PVDF–0.20WC has demonstrated the highest Σ<sub>R</sub> value, indicating effective shielding against fast neutrons, while PVDF–0.20WS2 had the lowest Σ<sub>R</sub> value. The Monte Carlo N-Particle Transport version 5 (MCNP5) simulations showed that PVDF–xWC attenuates gamma radiation more than pure PVDF, significantly decreasing the dose equivalent rate.Conclusion: Overall, this research provides insights into the radiation shielding properties of PVDF mixtures, with PVDF–xWC showing the most promising results.

Time-Lapse Pulsed-Neutron Logs for Carbon Capture and Sequestration: Practical Learnings and Key Insights

Pulsed-neutron logs are a staple of time-lapse monitoring programs for saline aquifer carbon capture and sequestration (CCS) projects and are unsurprisingly the most frequently run wireline logs in both injection and monitoring wells. While the emphasis imposed by government regulators and the focus of operators to date has been on the verification of CO2 containment, it is envisioned that a savvy interpretation of the multiple independent measurements should be able to unlock much greater value for the project than merely detecting the location of stored CO2. Recently introduced capabilities for novel measurements and improved environmental compensation should further increase the repeatability, interpretability, and value of these logs. We reviewed more than 30 time-lapse runs of pulsed-neutron logs acquired over a period of 15 years on three mature CCS projects using both previous- and new-generation pulsed-neutron tools, including measurements of formation sigma, hydrogen index, and fast neutron cross section. Special attention in processing is required when changes occur to the wellbore environment between runs, although this is mitigated by the improved environmental compensation scheme of the newer tool. We performed both standalone estimates of CO2 saturation from single-physics time-lapse measurements and simultaneous interpretation of multiple independent time-lapse measurements and studied the results side-by-side with openhole log interpretation, core analysis, and well test results from the evaluation phase. The apparent changes in saturation were framed within the context of the injection history and important events in the life of the wells. A first finding is that differences in apparent CO2 saturation between the various independent measurement physics of the pulsed-neutron tool are often reconcilable and may carry additional information about the state of the well or reservoir. With respect to verification of containment, depending on the well configuration, it may be possible to differentiate between CO2 in the formation and CO2 in the annulus. The interpreted CO2 saturation itself can have different significance depending on the timing of acquisition and the type of well. Measured at the right time, it is a direct in-situ measurement of formation CO2 storage efficiency. In other cases, the interpretation reveals formation dryout in the near-wellbore region of injection wells, a condition that may presage loss of injectivity. We now understand that it is important for operators to plan the timing and frequency of pulsed-neutron runs according to what they want to measure and not based solely on regulatory obligations. In a CCS project, time-lapse pulsed-neutron logs should be thought of as much more than simple indicators of the presence and migration of CO2. They give important information about migration pathways. They can also help to quantify essential uncertainties on reservoir performance that are difficult to ascertain during evaluation. For example, storage efficiency is difficult to quantify with openhole logs since the formation is initially at zero CO2 saturation. Yet it is one of the keys to determining the ultimate storage capacity of any reservoir. Time-lapse pulsed-neutron logs provide an abundance of information that, when properly history matched, can greatly improve our models of CCS reservoirs to better navigate both the economic and operational risks associated with these projects.

Neutron attenuation features and elastic properties of silicate glasses containing Ta2O5, and Li2O

In this investigation, the elastic properties and neutrons attenuation factors for some optical glasses containing Ta2O5, SiO2, and Li2O were reported. The present glasses were also consisted of ZrO2 and Nb2O5 in very small concentrations. The glasses are chemically defined as 26.47Li2O-5.88ZrO2-(20-x)Ta2O5-xNb2O5-47.06SiO2, where, x takes the values: 0, 2.94, 5.88, and 11.77 mol%. The elastic properties of these glassy specimens were determined by employing Makishima-Mackenzie's theory (M.M.T). By using the same method, moreover, the micro-hardness and Poisson's ratio were assessed. Cross sections for slow, moderated, and fissile neutrons were computed through standard expressions and models. In addition, the influence of the partial replacement of Ta2O5 by Nb2O5 on the parameters were also analysed. The glass with the lowest Nb2O5 content presented the highest cross sections for fast, moderated, and slow neutrons. The neutron-absorption ability of included glasses declined as glass density declined and Nb2O5 molar concentration increased in the glasses. Therefore, the sample with the highest Ta2O5 content is recommended for neutron absorption applications.

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

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

Recycling of waste cathode-ray tube glasses as building materials for shielding structures in medical and nuclear facilities

In view of the low market demand and toxic elements contained in cathode ray tubes (CRT), a sustainable method of recycling CRT glasses has become important from the perspectives of waste management and the ecosystem. This research investigated the environmental and economic benefits of the panel (PCRT) and cone (CCRT) glasses, from conventional CRT, as building materials in medical and nuclear facilities. The chemical composition of the two glasses was quantified using X-ray fluorescence (XRF) equipment. The nuclear shielding ability of the CCRT and PCRT was investigated by estimating interaction parameters directly linked to photons, charged particles (CP) (β, H+, He2+, and C6+), fissile neutrons, and thermal neutrons. The mass attenuation coefficients (MAC) of the glasses were determined by FLUKA simulations and XCOM. The CP stopping powers and range were computed using the ESTAR, PSTAR, ASTAR, and SRIM software. The cross sections for fast and thermal neutrons were used to determine the neutron moderating and absorption abilities of the two glasses. The CCRT showed greater gamma, fast neutron, thermal neutron, and CP attenuation capacity relative to the PCRT. The CCRT thus contributes more to the radiation shielding competence of CRT than the PCRT. The photon shielding proficiency of the glasses was found to be superior to many standard shielding materials. Hence, PCRT, CCRT, and therefore CRT could improve the shielding capacity of a medium. The use of CRT glass for producing radiation-protective glasses is a viable way of eradicating waste CRT from the environment.

Simultaneous evaluation of uranium and plutonium fast neutron fission cross sections up to 200 MeV for JENDL-5 and its updates

Simultaneous evaluation of 233,235,238U and 239,240,241Pu fission cross sections for fast neutrons up to 200 MeV was performed for the JENDL-5 library. Experimental covariances were estimated for each experimental dataset of cross sections or cross section ratios extracted from the EXFOR library, and they were stored in an experimental database dedicated to the new evaluation. The cross sections were expressed by Schmittroth’s roof functions, and the values on defined incident energy grids were adjusted by the least-squares method to reproduce the experimental cross sections and their ratios. The newly evaluated cross sections were validated using the spectrum averaged cross sections measured in the 252Cf spontaneous fission neutron standard field. The evaluation adopted by the JENDL-5 library was further updated by addition of two datasets and deletion of one dataset.

Multicomponent Ge-As-Te-Pb chalcogenide glasses for radiations shielding applications

Phy-X/PSD online program is used to obtain various radiation shielding indices in a photon energy region located between 0.15 and 15 MeV for Ge25-As10-Te65-x-Pbx (x = 2, 4, 6, 8, 10 at %) chalcogenide glasses. The linear attenuation coefficient LAC, mass attenuation coefficients MAC, effective atomic number Zeff, effective electron density Neff, half-value layer, HVL, tenth value layer TVL, mean free path MEF, energy absorption and exposure buildup factors (EABF, EBF), and fast neutron cross section FNRCS have been introduced. The findings conclude that the LAC and MAC measurements are greater and therefore better than commercial and traditional glasses. Also, it was found that HVL, TVL, and MFP were reduced with the addition of Pb to the tested glasses, which improve the shielding characteristics. Zeff and Neff of the compositions under study were varied as (42.58- 59.75) and (2.36-3.09 x 1023 electrons/g) respectively. A reduction was noticed in EBF and EABF values with the increment of Pb concentration in the investigated glasses at the entire photon energies, at all values of MFP that emphasize the enhancement of shielding properties of these glasses with the addition of lead. FNRCS were found to be changed between 0.87 and 0.095 Cm-1 as Pb content varies from 0.0 to 0.01 respectively that let these glasses

Silicate glass system and significant role of Bi2O3 on radiation protection ability against gamma, neutron, and charged particle

The use of ionizing radiation and hence, shielding materials has witnessed tremendous diversity in the last three decades. The use of glasses as shields in radiation facilities is gaining attention due to many attractive characteristics of different glass systems. In this study, the gamma-ray, neutron (fast, thermal, and cold), and charged particles’ (electron, proton, alpha particles, and carbon ion) shielding ability of 60SiO2–15ZnO– (25-x)Li2O–xBi2O3 glasses (x = 0 (SZL-00Bi), 5 (SZL-05Bi), 10 (SZL-10Bi), 15 (SZL-15Bi) and 20 (SZL-20Bi) mol%) was examined through the estimation and analysis of interaction parameters. The calculated shielding parameters showed strong dependence on radiation energy and glass composition. The mass attenuation coefficient of the glasses fluctuates between 0.0236 and 0.2223 cm2g-1 for SZL-00Bi, 0.0308–1.7083 cm2g-1 for SZL-05Bi, 0.0343–2.542 cm2g-1 for SZL-10Bi, 0.0360–3.0756 cm2g-1 for SZL-15Bi, and 0.0372–3.4463 cm2g-1 for SZL-20Bi. Also, for photons of energy within 0.1–10 MeV, the estimated mean free path in the SZL-xyBi glasses varies from 1.64 cm to 15.47 cm for SZL-00Bi, 0.17 cm–9.32 cm for SZL-05Bi, 0.10 cm–7.25 cm for SZL-10Bi, 0.07 cm–5.99 cm for SZL-15Bi, and 0.06 cm–5.33 cm for SZL-20Bi. Analysis of other gamma shielding parameters indicated that Bi2O3 enhanced the photon shielding ability of the glasses. The stopping power value of SZL-xyBi glasses for electron, proton, alpha particles and heavy carbon ion varies as follow: SZL-20Bi˂SZL-15Bi˂SZL-10Bi˂SZL-05Bi˂SZL-00Bi. In addition, SZL-00Bi has the highest RCSDA and RP whereas, the SZL-20Bi has the lowest estimated particle range among the studied glasses. The total removal cross section (ΣR) for fast neutrons is 0.0975 cm−1 for SZL-00Bi, 0.0991 cm−1 for SZL-05Bi, 0.0983 cm−1 for SZL-10Bi, 0.1013 cm−1 for SZL-15Bi and 0.1011 cm−1 for SZL-20Bi. The best thermal neutron absorber among the glasses is SZL-05Bi. A comparison of the photon shielding prowess of the present glasses with conventional shields showed the supremacy of the SZL-xyBi glasses. The present glasses thus have great potential as radiation shields.

Effect of the B4C content on microstructure, microhardness, corrosion, and neutron shielding properties of Al–B4C composites

Aluminum–Boron Carbide (Al–B4C) metal matrix composites (MMCs) are frequently studied because of their high thermal conductivity and ability to absorb neutrons, as an alternative material for the manufacture of baskets in spent fuel casks in the nuclear industry. Keeping in mind, this study aimed to produce Al–B4C MMCs and examine their properties to shield against neutron radiation. The Al-xB4C (x = 5, 10, 15, 20, 25, 30, 40, 45, and 50 wt%) composite powders were prepared by a high-energy planetary ball milling. The prepared powders were compacted with uniaxial cold compaction at 500 MPa and sintered in a tube furnace at 600 °C for 1 h under an Argon atmosphere. The microstructure of the obtained MMCs was characterized by using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). The XRD result indicates that the synthesized Al–B4C MMCs have a predominant phase of Al and B4C. The SEM images show the distribution of B4C reinforcement inside the Al matrix. The relative density, measured by the Archimedes principle decreased from 99.95 to 91.55% as the B4C ratio increased. The microhardness value was increased in the beginning from 44 to 75.5 HV and then decreased to 36.5 HV as the B4C ratio increased. The polarization curves show that the corrosion current density and the corrosion rate were significantly increased from 52.25 to 375.36 μA cm−2 and 0.573–4.121 mm/year as the B4C ratio increased. The simulation results, computed by the MCNP6.2 program show that the thermal neutron cross-section and the fast neutron cross-section was increased from 4.5 to 40.1 cm−1 and 0.13 to 0.15 cm−1 as the B4C ratio increased. Further, the neutron equivalent dose rate was calculated experimentally with the 241Am–Be neutron source and the data was completely agreed with the computational results.

Enhancing the physical, optical and shielding properties for ternary Sb2O3–B2O3–K2O glasses

Ternary Sb2O3–B2O3–K2O glass system, with general composition of x Sb2O3–(70-x) B2O3–30 K2O (where x = 0, 10, 20, 30, 40, 50) were prepared using melt-quenching technique. Structures of these glasses were investigated using XRD analysis and FT-IR spectroscopy. The optical properties were investigated using the UV–Vis NIR JASCO (Model V-670) Double Beam Spectrophotometer. Different physical parameters, such as density (ρ), molar volume (VM), oxygen molar volume (VO) and oxygen packing density values (OPD) have been estimated. Also, the Gamma radiation shielding ability have been characterized for the investigated glasses using Phy-X/PSD software in the photon energy range (0.015–15 MeV). XRD analysis confirm the amorphous nature of the prepared glasses. The optical energy gap of SBK glasses is decreasing from 4 to 2.63 eV with increasing Sb2O3 content while refractive index is increasing from 2.17 to 2.5 because of increasing the non-bridging oxygens (NBOs) in the glass matrix. The molar refractivity (Rm), molar polarizability (αm) and the third-order nonlinear optical susceptibility values χ3 have been calculated, their values are found to increase with increasing Sb2O3 content. Glass density and molar volume values for the SBK glasses increase with increasing Sb2O3 content. Increasing the Sb2O3 concentration enhance the radiation shielding features of the prepared glasses against gamma rays and neutrons. Hence, the mass attenuation coefficient (MAC) increases from 8.6 to 35.4 cm2/gm while the half value layer (HVL) decreases from 0.036 to 0.0046 cm at 0.015 MeV as the Sb2O3 concentration increases from 0 to 50 mol%. The fast neutron cross section of the six investigated SBK glasses are (0.087, 0.092, 0.095, 0.091, 0.094 and 0.092 cm−1), respectively.

Evaluation of the radiation shielding characteristics of WO3–MoO3–TeO2/Sb2O3 glasses

ABSTRACT This study reports the WO3–MoO3–TeO2/Sb2O3 glass system by means of its shielding ability against gamma photons, neutrons and charged particles. The studied glass system has two components, namely WO3–MoO3–Sb2O3 (denoted by WMS) and WO3–MoO3–TeO2 (denoted by WMT). The shielding properties of WMS and WMT glass systems were examined by using several theoretical methods and the FLUKA simulation code over energies between 0.1 to 10 MeV for gamma rays shielding application and between 0.01 and 10 meV for neutron shielding ability. The obtained results indicated that both linear attenuation coefficient and mass attenuation coefficient of WMS and WMT glasses are lowest at 6 MeV and they are highest at 0.10 MeV. The minimum half value layers (HVLs) of WMS and WMT glasses are at 0.10 MeV with values of 0.08506, 0.07822, 0.08268, and 0.07257 cm for WMS1, WMS2, WMT1 and WMT2, respectively. Maximum HVLs occured at 6 MeV with values of 3.89233, 3.70118, 3.89804, and 3.63977 cm for WMS1, WMS2, WMT1 and WMT2, respectively. The mean free paths of WMS and WMT glasses modify from 0.12272 to 5.61545 cm for WMS1, from 0.11284 to 4.33968 cm for WMS2, from 0.11928 to 5.62368 cm for WMT1 and from 0.10462 to 5.25107 cm for WMT2. The WMT2 has the largest values of effective electron densities between 0.10 and 0.6 MeV and WMT1 has the largest values between 0.8 and 10 MeV. The largest gamma dose rates are obtained at 10 MeV and the smallest values are obtained at 0.5 MeV. Both cross-sections for fast neutrons and (σ tot) for thermal neutrons change in order of WMS1 < WMS2 < WMT1 < WMT2.

Synthesis and properties of tellurite based glasses containing Na2O, BaO, and TiO2: Raman, UV and neutron/charged particle shielding assessments

In this paper, 75TeO2–5Na2O–20BaO‒xTiO2 (TNB-Tix with x = 0, 05, 10, and 15) glasses were synthesized by the conventional melt-quench technique using analytic grade tellurium oxide (TeO2), barium carbonate (BaCO3), sodium carbonate (Na2CO3) and titanium oxide (TiO2) as starting chemicals. The prepared glasses were studied for their physical features, Raman and UV spectra, and shielding performance against neutrons and charged particles. The optical property was investigated by UV–Vis spectrometry while the structural evolution of the glasses was studied through the Rahman spectra. Charged particles, slow and fast neutrons interaction parameters of the glasses were calculated theoretically and analyzed as well. The prepared glasses were yellowish without any flaws. The mass density of the glasses increased from 5.1 to 5.4 g/cm3 as TiO2 content declined from 15 to 0 mol%. Also, an improvement in the optical bandgap from 2.89 to 3.2 eV was recorded as BaO content increased concerning TiO2 while the refractive index declined from 2.43 to 2.35. Generally, the improvement in the TiO2 content of the glasses produced a rise in the total and scattering cross section of thermal and slow neutrons respectively. In addition, the fast neutron cross section was enhanced from 0.1005 to 0.1015 cm−1 for TNB-Ti00 – TNB-Ti15 glasses. The charged particle shielding parameters showed a strong dependence on the chemical structure of the glass system. The present glass system displayed good properties that could make them useful in optical and shielding applications.

A new method of gas reservoir evaluation based on neutron cross section logging

It is a widely used method to evaluate gas reservoirs by measuring inelastic gamma rays. A petrophysical parameter fast neutron cross section (FNXS) was proposed to describe the inelastic gamma for gas reservoir evaluation. FNXS is the 14 MeV neutron macroscopic elastic scattering cross section. However, the description of inelastic gamma using FNXS is only a first order approximation. Therefore, in some scenarios, the evaluation of gas saturation is not accurate enough. In this paper, we propose a new parameter total fast neutron cross section (TFNXS) based on the inelastic gamma generation and attenuation theory. TFNXS is the sum of macroscopic elastic and inelastic scattering cross sections of 14 MeV neutrons. The physical meaning of TFNXS is the probability of inelastic and elastic scattering between neutrons and formation. TFNXS is a new formation property independent of neutron macroscopic capture cross section (Sigma). TFNXS value of gas is quite different from that of matrix minerals, water, oil and clay, so TFNXS is more sensitive to gas-bearing formation and can identify gas reservoir more effectively. We compared the sensitivity of TFNXS and FNXS to gas. The sensitivity of TFNXS to gas is similar to that of FNXS. The effects of different formation conditions on the measurement accuracy of the TFNXS and the FNXS are analyzed with the Monte Carlo method. Compared with FNXS, TFNXS can be obtained more accurately in water and oil layers. Different clay minerals have different effects on TFNXS and FNXS. TFNXS is less affected by wellbore environment, such as well diameter and wellbore fluid, than that of FNXS. We evaluate the application results of TFNXS and FNXS in various model wells. Both TFNXS and FNXS can effectively identify gas-bearing formations. The accuracy of using TFNXS to obtain gas saturation is higher than that of FNXS. TFNXS can more accurately monitor reservoirs and evaluate low porosity gas reservoirs than FNXS.

Mechanical and radiation shielding properties of concrete containing commercial boron carbide powder

Boron-containing compounds have an excellent radiation shielding capability against slow neutrons. Moreover, sassolite in boron-containing compounds adversely affects cement hydration by retarding cement setting times. Accordingly, experimental and analytical investigations were carried out to examine the filler effect of commercial boron carbide (B4C) fine powder with a mean particle diameter of 2.64 μm on Portland cement hydration reaction, and concrete mechanical and radiation shielding properties. Isothermal calorimetry method was used to study the heat evolution and the cement degree of hydration of mixes containing 0, 25, 50 and 75% of B4C to cement by weight. Effects of adding B4C on concrete compressive strength were then investigated. Analytical investigations, specifically mass attenuation coefficient using WinXcom program and effective atomic number using Auto-Zeff, were carried out to evaluate the radiation shielding of corresponding concrete mixes for photon energies ranging from 10 keV to 10 MeV. Moreover, macroscopic effective removal cross-sections for fast neutrons and thermal neutron macroscopic absorption cross-sections were calculated using MRCsC program and JANIS-4 software, respectively. The results confirm that the presence of sassolite in the studied B4C powder retards the hydration reaction for the first 24 h and reveal significant improvements in concrete strength thereafter with the increase of boron carbide due to the filler effect. Furthermore, the addition of boron carbide yielded measurable improvement in neutron shielding capabilities of concrete mixes.