Total Macroscopic Cross Section Research Articles

A study on the interaction parameters of charged and uncharged radiation types with some indoor plants

This study investigates how gamma rays, neutrons, and electrons interact with five commonly found indoor plants: Spathiphyllum wallisii (SW), Ficus elastica (FE), Dieffenbachia camilla (DC), Schefflera arboricola (SA), and Ficus benjamina (FB). Utilizing experimental measurements (with HPGe detector), Monte Carlo simulations (GEANT4 and FLUKA), and theoretical calculations (ESTAR and WinXCOM), some radiation interaction parameters for gamma rays, fast neutrons, thermal neutrons, and electrons were determined. Secondary particle generation was also analyzed to provide a comprehensive assessment. The determined linear attenuation coefficients with the help of the WinXCOM are 0.1376, 0.1662, 0.1385, 0.1651 and 0.1698 cm-1 for SW, FE, DC, SA and FB, respectively. The calculated total macroscopic cross sections for indoor plants in the same sample order are 2.0290, 2.0350, 2.0285, 2.0363 and 2.0362 cm-1. Among the investigated plants, FB exhibited the highest gamma ray interaction, while SA and FB showed superior interaction against fast neutrons compared to SW and DC. The findings reveal significant variations in interaction effectiveness and secondary radiation production across these plants, offering valuable insights for radiation safety and environmental health evaluations.

Uniformity of neutron absorber distribution in Gd-containing neutron absorber materials

For neutron absorber materials (NAMs), it is important to evaluate their neutron-absorption capability and quantitatively evaluate the uniformity of the distribution of neutron-absorbing elements within the materials. In this study, neutron attenuation testing with cold neutrons was conducted on Ti-based model alloys with Gd concentrations ranging from 1 to 10 wt%. The neutron attenuation coefficients (μ) of the Ti-Gd alloys were determined and compared with the total macroscopic cross sections (Σt), which were computed using the latest ENDF/B library. Although the measured μ increases with the Gd-content, the coefficients were not linearly proportional to the absorber amount, and the ratio of μ to Σt gradually decreased. This was caused by beam hardening, which is more significant for higher-absorbing and thicker NAMs. To examine the uniformity of the neutron absorbers (NAs) in plates, we propose a method for characterizing the degree of homogeneity by introducing a radial distribution function (RDF). After the optical microscopic images of the Ti-Gd alloys were converted to grayscale, the RDF of the Gd particles in the alloys was calculated under the assumption that a large particle is composed of many adjacent unit particles. A unit particle was regarded as one pixel in the digital image. For all Ti-Gd alloys, the RDF had a value of > 1 at short interparticle distances, which indicated a dense population of particles in a small area due to cluster formation. The method of using the RDF as a uniformity parameter was validated through a case study involving a fixed NA content with different distributions.

A deep-learning representation of multi-group cross sections in lattice calculations

To compute few-group nodal cross sections, lattice codes must first generate multi-group cross sections using continuous energy cross-section libraries for each material in each fuel cell. Since the processing cost is significant, we propose representing the multi-group cross sections during lattice calculations using a pre-trained deep-learning-based model. The model utilizes a combination of Principal Component Analysis (PCA) and fully connected Neural Networks (NN). The model is specifically designed to manage extensive multi-group cross-section data sets, which contain data for several dozen nuclides and encompass more than 50 energy groups. Our testing of the trained model on a PWR assembly with a realistic boron letdown curve revealed an average relative error of around 0.1% for both fission and total macroscopic cross sections. The average time required for the model to generate the cross sections was approximately 0.01% of the time needed to execute the cross-section processing module.

Ulexite/HDPE-Bi2O3/HDPE layered composites for neutron and gamma radiation shielding

In this study, 5, 10, and 15 wt% ulexite (NaCaB5O6(ΟH)6·5H2O, hydrated sodium calcium borate hydroxide) and 15 wt %Bi2O3 filled high density polyethylene (HDPE) composite materials were fabricated through conventional melt-extrusion processing techniques in the form of a layered structure in order to absorb both neutron radiations and the secondary radiations resulting from neutron induced reactions. In the layered structure, HDPE was used to slow down neutrons, while ulexite and Bi2O3 were used to capture thermal neutrons and secondary gamma radiations, respectively. The properties of ulexite/HDPE and Bi2O3/HDPE composites were investigated through scanning electron microscopy (SEM), X-ray diffraction (XRD), thermogravimetric analysis (TGA). The mechanical properties (tensile and hardness) of composite materials were also investigated. The results showed that the addition of ulexite and Bi2O3 particles does not change the thermal and mechanical properties of the HDPE composites significantly. Total macroscopic cross-section of the composites was determined using a 239Pu–Be (α,n) neutron source, while their linear and mass attenuation coefficients were determined using a137Cs gamma-ray source. The results show that ulexite filled HDPE composites enhance neutron shielding property and Bi2O3 filled HDPE composites provide good shielding performance for gamma rays.

Eco-friendly transparent glass prepared from rice straw ash for neutron and charged particle radiation shielding

The current study aimed to fabricate and investigate new glass system incorporating silica from rice straw ash (RSA). The effect of adding different concentrations of bismuth oxide (Bi2O3) on the glass's structure was examined using FTIR. The A.C. conductivity of the prepared glasses over the frequency range (120 Hz − 100 kHz) and the temperature range (300–620 oK) have been measured. Based on its electron density, a substance's electrical conductivity can alter as photon energy increases. For the current investigated glasses, the electron density (Neff) and electrical conductivity relation over the energy range (0.015–10 MeV) were calculated. Glass samples' ability to attenuate radiation such as neutrons, electrons, protons, and alpha particles has been investigated. For neutrons, the effective removal cross-section (∑R, cm−1) and the total macroscopic cross-section (∑T, cm−1) are calculated. For charged particles like protons (H+) and alpha (He2+), SRIM software has been used to determine the glass Mass Stopping Power (MSP) and the Projected Range (PR). For electrons (β−), MSP and the Continuous Stopping Down Approximation (CSPA) were calculated using ESTAR (NIST). On the other hand, Phy X/Zextra software was used to determine glasses effective atomic numbers (Zeff) over a broad energy range. As a result, BSiBi20 glass with the highest Bi2O3 concentrations (20 mol. %) shows superior shielding features for fast neutrons (∑R = 0.128 cm−1) compared to the competitors like ordinary concrete (O.C.). For charged particles, BSiBi20 glass achieves the lowest MSP, CSPA, and the highest Zeff values over the entire energy range. The KERMA (Kinetic Energy Released per unit Mass of the Absorbing material) calculations indicated a significant influence on the elemental composition of the glass system in the intermediate photon energy regime.

A detailed investigation of gamma and neutron shielding capabilities of ternary composites doped with polyacrylonitrile and gadolinium (III) sulfate

The shielding efficiencies of gamma and neutron radiations for ternary composites containing polyester resin, polyacrylonitrile and gadolinium (III) sulfate at different ratios were investigated in the present study. In order to investigate the gamma radiation shielding capacity of the produced ternary composites, linear and mass attenuation coefficients, half value layer, effective atomic number and radiation protection efficiency parameters were determined experimentally, theoretically and using the GEANT4 simulation code. The gamma shielding capabilities of the composites were studied in the photon energy range of 59.5–1332.5 keV. In order to investigate the neutron shielding abilities of composites, inelastic, elastic, capture and transport numbers, total macroscopic cross section and mean free path parameters were determined with the help of GEANT4 simulation code. In addition, the number of transmitted neutrons at different sample thicknesses and neutron energies were also determined. It was observed that gamma radiation shielding properties were improved due to the increasing amount of gadolinium (III) sulfate and neutron shielding properties were improved due to the increasing amount of polyacrylonitrile. While the composite coded P0Gd50 exhibits a better gamma radiation shielding ability than the others, the neutron shielding of the sample coded P50Gd0 is also more favorable than the others.

Monte Carlo simulation study on Barite Concrete for neutron-gamma photon

In the ongoing research, the Monte Carlo simulation of three different Barite Concrete samples coded as C-B0, C-B50 and C-B100 was estimated via MCNPX Monte Carlo Code to investigate the neutron-gamma photon shielding characteristics. Simulations were performed for 252Cf source which is surrounded by the polyethylene cylinder. Using Watt Fission (WF) distribution and Doppler Effect (DE), the neutron-gamma photon flux was obtained and was compared to various types of Barite Concrete samples. The total macroscopic cross section, linear attenuation coefficient (µ), Half Value Thickness (HVT) and Mean Free Path (MFP) were calculated for gamma rays and neutrons. The equivalent absorbed dose rate was estimated for neutrons and showed that an increment of the Barite rate causes a reduction in the equivalent absorbed dose rate from 0.0003572 to 0.0003542 (µSV/h). Moreover, total macroscopic cross section which shows total macroscopic cross section which shows the neutron attenuation coefficient (∑t ) undergoes descending trend of (∑t )C-B0 < (∑t )C-B50 < (∑t )C-B100 which confirms the disability of the Barite Concrete sample against neutrons. The evaluated gamma photon attenuation parameters for energy ranging from 0.05 to 11 MeV are shown graphically. Findings reveal that by increasing the rate of the Barite, the samples’ shielding ability will reduce against neutron whereas it will enhance the gamma photons shielding capability.

Optical, elastic, and neutron shielding studies of Nb2O5 varied Dy3+ doped barium-borate glasses

A sequence of modifier oxide (Nb2O5) varied Dy3+ doped barium-borate glasses have been manufactured. Various structural and optical characterizations are done. Investigation of the local symmetry around the Dy3+ in the host medium is predicted through the bonding parameter analysis and Judd-Ofelt intensity parameters determined by employing the least-squares fitting method. Various radiative properties for the emission transitions 4F9/2→6H15/2, 4F9/2→6H13/2, and 4F9/2→6H11/2 of the Dy3+ are estimated. Further, the Y/B ratio intensity ratio values, CIE color coordinates, color purity, and color-correlated temperature of the prepared glasses were calculated using the luminescence spectra. Based on the spectroscopic quality factor and CIE coordinates, the prepared glasses expose their aptness for the laser media and white-light-emitting devices. Neutron shielding efficiency is evaluated by the computation of fast neutron effective removal cross-section and neutron scattering, absorption, and total macroscopic cross-sections. It was found that modification by Nb2O5 makes an enhancement mainly in fast neutron-shielding competence, but it also has some minor influence on thermal neutron absorption efficiency.

Monte Carlo Simülasyonu Kullanılarak B4C, B2O3, Sm2O3 ve Gd2O3 Katkılı Polimer Matrisli Kompozitlerin Termal ve Hızlı Nötron Zırhlama Özelliklerinin İncelenmesi

In this study, thermal (2.53*10-8 MeV) and fast (2 MeV) neutron total macroscopic cross-sections of paraffin, polycarbonate, and polyester matrix polymers doped with B4C, B2O3, Sm2O3, and Gd2O3 (at weight percentages of 5%, 10%, 15%, 20%, and 25%) were computed by using Monte Carlo simulations. Additionally, the macroscopic effective removal cross-section ) of fast neutrons was theoretically computed based on the mass removal cross-section values ) for various elements in polymers and additives. The obtained results show that the highest thermal neutron total macroscopic cross-section was obtained in polycarbonate doped with Gd2O3, and the highest fast neutron total macroscopic cross-section was observed in paraffin doped with Sm2O3. Besides, the paraffin provided the highest fast neutron total macroscopic cross-section for all additives. The results of this study provide a good understanding of shielding properties of paraffin, polycarbonate, and polyester matrix polymers doped with B4C, B2O3, Sm2O3, and Gd2O3 against thermal and fast neutrons.

Monte Carlo simulations study on gamma ray–neutron shielding characteristics for vinyl ester composites

AbstractThe current study aims to report the gamma ray–neutron attenuation parameters of eight types of vinyl ester pumice (V‐P) composites for wide energy ranges of 511–2140 keV via MCNPX and FLUKA codes. The radiation shielding properties of these composites were presented graphically and are discussed in detail in this study. From the obtained results, it can be concluded that the MCNPX and FLUKA outcomes are in agreement with each other, and also the FLUKA code can be used as an alternative technique to determine the attenuation features of vinyl ester composites. Besides, the linear attenuation coefficient and total macroscopic cross sections' (∑t) dependencies on the density and also mass attenuation coefficient were found for all selected samples. To sum up, findings show that composite coded as V‐P50 can be a good competitive candidate in terms of gamma rays and neutron shielding.

Neutron and gamma radiation shielding Ni based new type super alloys development and production by Monte Carlo Simulation technique

In the present study, five different type of new super alloys have been developed and produced which have high temperature resistance for both gamma and fast neutron radiation shielding. These super alloys have been produced with powder methodology by using the materials such as iron (Fe), rhenium (Re), boron carbide (B4C), nickel (Ni), chromium (Cr), boron (B), copper (Cu), tungsten (W), tantalum (Ta). Gamma-ray Mass attenuation coefficients, half value layer effective atom number, mean free path and fast neutron total macroscopic cross section, transmission number values have been theoretically calculated by using Geant 4 code, WinXCom and Phy-X / PSD software. In addition, experimentally equivalent dose measurements have been carried out by using 241Am-Be fast neutron source. Sulfuric acid abrasion, compressive strength, temperature resistance and weldability tests have been made. The results are compared with 316LN nuclear stainless steel because of its common use in nuclear shielding applications and all new type super alloy samples have shown better radiation absorption ability for both gamma and fast neutron. It has been determined that these new type super alloys can be used in nuclear applications for better shielding.

Gamma / neutron radiation shielding, structural and physical characteristics of iron slag nanopowder

The current study explores the effectiveness of an iron slag nanopowder (ISNP), which prepared from local iron steel industry, against gamma/neutron shielding. The structural and physical characteristics were experimentally determined. The crystal structure, morphology and elemental composition of the ISNP were studied using X-ray diffraction, scanning electron microscopy (SEM), energy dispersive X-ray analysis (EDX). On the other hand, a comprehensive experimental and theoretical study on the γ-ray shielding effectiveness of the ISNP was performed. In which, the experimental results have been validated by comparison with theoretical data which obtained by using the WinXcom program. This paper uses the well-known mathematical relationships to derive many shielding and dosimetry parameters such as effective atomic number and, effective electron density, for photon interaction and photon energy absorption as well as the γ-ray kerma coefficient from the mass attenuation coefficient. The exposure and energy absorption buildup factors have been also calculated. Furthermore, the shielding effectiveness against thermal and fast neutrons has been tested in terms of total macroscopic cross-sections. The results have revealed that the produced ISNP, with crystallite size of 24.5nm, exhibits good shielding characteristics. Finally, based on this preliminary study, we can have concluded that the iron slag nanoparticles can be suitably used as an effective and safe (lead - free) component for radiation shielding.

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.

Development and production of metal oxide doped glasses for gamma ray and fast neutron shielding

In this study, new glasses were designed and produced that did not consist of toxic lead oxide (PbO). Unlike conventional glass with powdered silicon oxide (SiO2) added [cobalt oxide (CoO), cadmium tungsten oxide (CdWO4), bismuth oxide (Bi2O3), chromium oxide (Cr2O3), and boron oxide (B2O3)], new glass specimens were fabricated via melt-quenching. The total macroscopic cross-sections, mean free path, and transmission number were calculated using GEANT4 Monte Carlo code. The equivalent dose rates were measured for 4.5 MeV fast neutrons. Gamma attenuation experiments were conducted at several energies in a range of 53–383 keV. The experimental and theoretical results were compared with those of construction concrete, paraffin, glasses containing lead, borosilicate, normal window, and soda glasses. The new glass minimized both gamma and neutron radiation leaks that may occur in various fields such as radiation treatment, nuclear power plants, radioactive waste transport and storage, and space and laboratory research.

MoO3 reinforced Ultra high molecular weight PE for neutrons shielding applications

Structural, Morphological and neutron attenuation properties of ultrahigh molecular weight polyethylene (UHMWPE) reinforced with molybdenum trioxide composite material obtained by solid state mixing of pellets and powder components via hot pressing have been studied. In this work, Ultra high molecular weight polyethylene and MoO3 has been utilized with different concentration using a hot compression iron mold. The changes in the microstructure of the prepared polymer and its particulate composites were evaluated by SEM. In order to investigate the quantitative of all elements used in these samples, we utilized Energy-dispersive X-ray spectroscopy (EDX) to present the profiles of synthesized samples which show a clear peaks C (carbon), and Mo (Molybdenum). The X-ray diffraction of all samples was examined to study the modification of the phase and lattice for UHMWPE before and after the addition of molybdenum trioxide (MoO3). Raman analysis was utilized to explore the crystallinity, molecular interactions, phase, polymorphy, and chemical structure. This technique is based on non-destructive interaction between light with the chemical bonds inside a material. The Raman spectrum for the current work presented within the range of 150–1600 cm−1. In order to understand the neutron shielding performance for the fabricated UHMWPE with different concentrations of MoO3, 241Am/Be fast neutron source was used to determine the fast neutron radiation equivalent dose rate. The background absorbed dose rate was measured as 0.98 (±0.017) (μSv/h). It was seen that the addition of MoO3 into the UHMWPE enhances the neutron shielding. The dose released from the source was absorbed by 7.71 (±1.3) %, 18.48 (±2.9) %, 22.26 (±2.4) %, 24.07 (±1.8) % and 26.17 (±1.3) % from UHMWPE with 1, 2, 3, 4 and 5% of MoO3 respectively. Additionally, GEANT4 10.02. p01 version was used to calculate total macroscopic cross-sections (TMCS) for the MoO3 doped UHMWPE. According to the results obtained from GEANT4, it was observed that the TMCS values of the UHMWPE increase with increment of the mole fraction of the MoO3 in the structure of UHMWPE examined and that the 5.0% MoO3 sample has the best neutron shielding ability.”

A wide energy range neutron shielding material based on natural rubber and boron

Neutron have been used in many applications. Neutrons have no electric charge and have high penetrating power. Therefore, the use of neutron shielding material is one of the methods to effectively prevent neutron radiation hazard. In this work, neutron shielding materials were designed using the Monte Carlo N-Particle (MCNP) transport code in order to select proper materials for a wide energy range neutron shielding. Thicknesses of 1 cm and 4 cm were chosen for shielding neutron in the wide energy spectrum (10E-8 - 100 MeV). MCNP simulation results indicated that the best shielding material was 2 layers of 2 cm natural rubber (NR) and 2 cm NR with 20 part per hundred rubber (phr) boron powder. This shielding material can reduce the neutron dose by 17.91 ± 0.09 % with the total macroscopic cross section of 0.042 cm−1. The optimal neutron shielding materials were fabricated. The mechanical properties and aging properties of shielding materials was found to meet the requirement of Thai Industrial Standard (TIS 2377-2551 Rubber flooring). Therefore, the designed shielding material can be utilized to shield against wide energy range neutrons.

Epoksi Bazlı Metal Ve Metal Oksit Katkılı Yeni Kompozit Nötron Ve Gamma Radyasyon Moderatör Malzemesi

In this study, five different composite shield materials were developed and produced. Epoxy resin based different proportions, materials such as [chromium oxide (Cr2O3), lithium florür (LiF), nickel oxide (NiO), bismuth oxide (Bi2O3), manganese oxide (MnO), copper oxide (CuO), titanium oxide (TiO2), cobalt oxide (CoO), gadolinium oxide (Gd2O3), selit (CaWO4), iron oxide (Fe2O3), lead oxide (PbO)] were used on production. The GEANT4 code of the Monte Carlo simulation program was used in determining mixing ratios. The total macroscopic cross section, mean free path and transmission neutron number were determined for fast neutron radiation by using GEANT4 simulation code. The mass attenuation coefficient and half-value layer (HVL) were calculated for gamma radiation by using WinXCom software. Experimental absorbed dose measurement was carried out and in these measurements 241Am-Be neutron source with 74 GBq activity which average neutron energy is approximately 4.5 MeV and BF3 gas detector were used. Both simulation and experimental measurements were compared with paraffin, conventional concrete. It was found that the new composite shielding material absorbed radiation much better than these reference materials. It has been suggested that this new radiation shielding composite material can be used in areas such as nuclear medicine, transport and storage of radioactive waste, nuclear power plants and as shield material for both neutron and gamma radiation.

High alloyed new stainless steel shielding material for gamma and fast neutron radiation

Abstract Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

Comparison of beta and neutron attenuations of poly(imide siloxane) block copolymers for medical applications

Beta and neutron attenuations of poly(imide siloxane) block copolymers were compared to evaluate radiation penetrations into flexible sheet (BI) and pellet form BII grains. The modification of BII polymer has improved its radiation shielding properties of poly(imide siloxane) block copolymer as the result of the pelletization. The linear attenuation coefficient of BII sample was determined as ∼0.073 (1/cm) and the half value layer was at ∼9.5 cm for negative beta particles. Besides, the total macroscopic cross section was at 0.047 (1/cm) and the half value layer of BII sample was determined at 14.74 cm for neutrons. The variations of beta and neutron attenuations indicated the structural changes in the poly(imide siloxane) block copolymer depended on the critical production parameters at the production step. The pelletization of the grains of BII improved beta and neutron shielding properties of poly(imide siloxane) block copolymers for the use in medical applications.

New high temperature resistant heavy concretes for fast neutron and gamma radiation shielding

Abstract In the present work, we developed three new high temperature resistant heavy concretes as novel radiation shielding materials. For this purpose, chrome ore (FeCr2O4), hematite (Fe2O3), titanium oxide (TiO2), aluminum oxide (Al2O3), limonite [FeO (OH) nH2O], siderite (FeCO3), barite (BaSO4), nickel oxide (NiO) materials and alumina cement were used. Mass combination ratios of components and total macroscopic cross sections (scattering, absorption, capture, fission) of the samples were calculated by using GEANT4 code. The resistances of the prepared samples were evaluated in terms of compression strength after exposure at the 1000 °C temperature. Neutron equivalent dose rate measurements were carried out by using 4.5 MeV 241Am-Be neutron source and BF3 detector. All results were compared with normal weight concrete and paraffin. The results of neutron dose indicate that neutron absorption ability of the new heavy concretes is higher than normal weight concrete and paraffin. In addition to neutron measurements, different γ-ray shielding parameters such as mass attenuation coefficient (MAC), effective atom numbers (Zeff), half value layer (HVL) and mean free path (MFP) have been calculated using WinXCOM software in order to investigate the effectiveness of using the prepared concretes as a radiation shielding materials. Gamma-ray results were compared with concretes and Pb-based glass.