Gamma Radiation Shielding Research Articles

Impact of micro/nano cadmium oxide on shielding properties of cement–ball clay matrix

This study investigates the gamma radiation shielding properties of cement–ball clay matrix composites doped with micro- and nano-sized cadmium oxide (CdO) particles. The linear attenuation coefficient (LAC) was determined using a sodium iodide (NaI) detector and five radioactive point sources with energies ranging from 59.5 to 1408 keV. The LAC values obtained were compared to the XCOM database and found to be in good agreement. The composites' half-value layer (HVL), tenth value layer (TVL), mean free path (MFP), effective atomic number (Zeff), equivalent atomic number (Zeq), and absorption buildup factor (EABF) were determined. The results showed that the addition of CdO particles improved the radiation-shielding behavior of the composites and increasing the weight fraction of CdO particles increased the shielding effectiveness. The results also illustrated that when nano-sized CdO particles were compared to their micro-sized counterparts, there was a significant enhancement in radiation shielding effectiveness. For instance, a composite material composed of 50% cement, 41.7% ball clay, and 3.8% nano CdO at an energy level of 0.0595 MeV exhibited a remarkable 12.2% increase in attenuation, surpassing the performance of the micro-sized sample with an equivalent concentration. Similarly, another composite consisting of 50% cement, 33.3% ball clay, and 16.7% nano CdO demonstrated a significant 15.4% increase in attenuation at the same energy level, when compared to the micro-sized sample. The study demonstrates the potential of CdO-doped cement–ball clay matrix composites for gamma radiation shielding applications.

Impact of PbO on structural, thermal, optical and gamma ray shielding behaviour of TeO2-CdO based glasses

A study was conducted to investigate the effect of an increase in the concentration of PbO on the structural, thermal, optical, and gamma ray shielding behavior of TeO2-CdO based glasses. The glasses were found to contain TeO4 and TeO3 units, as well as CdO metal oxide clusters, as indicated by the Raman spectrum. The glasses displayed minor weight loss (approximately 3%), suggesting high resistance to thermal degradation. With an increase in the concentration of PbO, the glass transition temperature (Tg), onset temperature (To), and crystallization temperature (Tc) all decreased. Additionally, the indirect band gap energy decreased from 3.497–3.337 eV as the lead oxide concentration increased. Moreover, the effects of PbO concentration on the linear attenuation coefficient (LAC) were investigated. The LAC values were found to be highest at 39.5 keV and decreased with increasing energy. As the PbO concentration increased, LAC values also increase. The Zeff displayed an inverse relationship with energy, except for an abrupt increase at 121.8 keV. The half value layer was found to be positively correlated with energy and varied with changes in PbO and CdO concentrations. The glasses demonstrated effective gamma radiation shielding properties, with the HVL remaining small even at higher energies, suggesting potential applications in industries such as medical imaging where thin shielding materials are preferred.

Gamma shielding performance of B2O3/BaO-based glassy system: Synthesis and simulation study

In this paper, the synthesis and gamma-radiation shielding attributes of xMgO-(30-x)BaO-69.5B2O3-0.5Fe2O3 glasses are reported. The aim is to ascertain the optimum chemical structure of the glass system for gamma radiation shielding applications and dosimetry. The simple melt and quench method was adopted for the synthesis of the xMgO-(30-x)BaO-69.5B2O3-0.5Fe2O3 glasses, where for x = 0, 5, 10, 12, and 15 mol% represented MBBFe-0, MBBFe-1, MBBFe-2, MBBFe-3, and MBBFe-4, respectively. The mass attenuation coefficients μ/ρ of the glasses were evaluated using the XCOM and FLUKA simulations for the photon energy range of 15 keV–15 MeV. The differences between the values of μ/ρ evaluated with FLUKA and XCOM were within 0.96%. The mass attenuation coefficients of MBBFe-0, MBBFe-1, MBBFe-2, MBBFe-3, and MBBFe-4 ranged from 0.0288 to 28.6205 cm2/g, 0.0279–25.6462 cm2/g, 0.0268–22.2712 cm2/g, 0.0260–20.4073 cm2/g and 0.0251–18.4086 cm2/g, respectively. The corresponding values of the half-value layer at 15 MeV are 7.06 cm, 7.64 cm, 8.25 cm, 8.70 cm, and 9.53 cm for MBBFe-0, MBBFe-1, MBBFe-2, MBBFe-3, and MBBFe-4, respectively. The specific gamma constant values of MBBFe-X glasses are in the range 11.289–264.417 Rm2/Cihr, 10.871–236.716 Rm2/Cihr, 10.386–205.282 Rm2/Cihr, 9.931–187.921 Rm2/Cihr and 9.444–175.515 Rm2/Cihr for MBBFe-0, MBBFe-1, MBBFe-2, MBBFe-3, and MBBFe-4, respectively. The substitution of BaO with MgO therefore, depreciated the gamma shielding efficacy, photon energy absorption abilities, and interaction per unit mass of the MBBFe-X glasses. This study concludes that MBBFe-0 is the best gamma absorber among the studied glasses and has the potential to be a good gamma-radiation shield with an efficiency that supersedes many traditional and modern shielding materials such as concrete, metal alloys, polymers, rocks, commercial glass shields, and other shielding glasses.

Experimental investigation of gamma radiation shielding ability of some building materials in Bitlis (Turkey)

Considering that people spend most of their time in buildings (home, office, school, shopping mall, etc.), it is seen that the radiation retention properties of building materials are important. Nuclear power plants are a complex of structures that are of great importance in today's energy-age world, but also bring some dangers. The most important of these dangers is the possibility of radiation leakage from these structures in the event of an explosion. In addition, shielding for radiation protection is important in medical applications and industry where radiation is widely used. Therefore, the strength and durability properties of concretes to be used in the construction of nuclear power plants need to be meticulously designed. All this shows how important and worthy of research the subject of the study is. For this purpose, new materials that are easy to access, abundant and cheap, and have high absorption capability are needed, rather than shield made only of lead materials. Perlite, pumice and Ahlat stones are abundant in Bitlis (Turkey) province and are extracted from the quarries. The bims block samples (B-1, B-2 and B-3) produced in factories in Bitlis. This study was carried out to investigate whether these materials, which are used as building materials, have radiation absorbing properties in terms of their use in different areas. Linear attenuation coefficients were calculated experimentally for gamma energies at 88.04 ± 0.01 keV, 356.50 keV, 661.60 keV, 834.80 ± 0.04 keV and 1173.20–1332.50 keV (for 109Cd, 133Ba, 137Cs, 54Mn and 60Co radioactive point sources) using gamma spectrometry with a NaI scintillation detector. In addition, photon absorption percentages, half value layer (HVL), tenth value layer (TVL) and mean free path (MFP) values of the samples were calculated. The obtained data were evaluated by comparing with similar studies in this field and the results obtained for lead shield. As a result of the study, it is seen that instead of using perlite and pumice alone, it is better radiation trap when they are mixed with each other and formed into bims blocks.

Effect of chromium substitution on the gamma and neutron radiation shielding properties of calcium hexaferrite nanoparticles

Effect of chromium substitution on the gamma and neutron radiation shielding properties of calcium hexaferrite nanoparticles

Computation of gamma-ray buildup factors for (25-x)CaF2-xAl2O3–15Bi2O3–59B2O3–CuO glasses

High energetic gamma rays are used in different fields and thus it may cause high risks for public if it is absorbed over limited dose. In order to be protected from these hazardous effect of radiation, the shielding is an effective ways and this requires shielding materials. For this purpose the gamma radiation shielding parameters for four different types glasses which was formulated as (25-x)CaF2-xAl2O3–15Bi2O3–59B2O3–CuO (x = 3,6,9,12 mol%) have been investigated. For this purpose, the linear absorption coefficients (LAC, μ) were calculated using Phy-X/PSD within the gamma ray energy range of 10−3-105 MeV. From using the equivalent atomic number (Zeq) and Geometric progression (G–P) method, the exposure buildup factor (EBF) and energy absorption buildup factor (EABF) values were evaluated at 0.015 MeV–15 MeV gamma rays. The values of EBF and EABF were calculated for 1, 5, 10, 20, 30, and 40 cm mean free path (mfp).

Boron containing polyvinyl alcohol/ polyethylene oxide/polyvinyl pyrrolidone composites: Preparation, characterization, gamma radiation shielding and gamma radiation effect on it's thermal properties

The production of PVA/PEO/PVP-B4C and PVA/PEO/PVP-BN nanocomposites is presented in this study by adding certain proportions of boron nitride (BN) and boron carbide (B4C) to the polyvinyl alcohol(PVA)/polyethylene Oxide (PEO)/polyvinyl pyrrolidone (PVP) blend. The percentages of BN and B4C nano powder in the PVA/PEO/PVP blend have been determined as 5%, 10% and 20%. Thermal characterizations were carried out with different techniques such as Differential scanning calorimetry (DSC), Differential Thermal Analysis (DTA), Thermogravimetry (TG). Surface, crystal structure and atomic percentage distribution analyzes of the obtained composites were performed by X-ray Photoelectron Spectroscopy (XPS), X-Ray Diffraction Analysis (XRD), Scanning Electron Microscopy-Energy Dispersive X-Ray (SEM-EDX), and elemental analysis. 10 kGy gamma irradiation was performed on PVA/PEO/PVP-B4C and PVA/PEO/PVP-BN composites. The thermal behavior of the composites at this dose level was investigated. In addition, the radiation shielding properties of composite films obtained at 10 cm distances using Am-241 beam source were examined.

Physico-mechanical properties and shielding efficiency in relation to mineralogical and geochemical compositions of Um Had granitoid, Central Eastern Desert, Egypt

The current work aims to describe the physico-mechanical characteristics and shielding efficiency with reference to the mineralogical and geochemical compositions of the Neoproterozoic Um Had composite granitoid pluton in order to deduce their favorability as dimension stones. The Um Had granitoid pluton has an elliptical outline with a mean diameter of about 10 km. This pluton is a composite (ranging from white to reddish pink color), hard, massive, and medium- to coarse-grained granitoid body. It is classified as syenogranite according to their modal and bulk chemical compositions. Geochemically, the granitoid pluton is a highly calc-alkaline, peraluminous granite, formed by low degree partial melting of tonalitic source rock in a post-collisional tectonic setting. The physico-mechanical properties of the granitoid pluton under study satisfy the requirements of dimension stone in terms of their bulk density (from 2561 to 2564 kg/m3), and to some extent water absorption capacity (from 0.38% to 0.55%). However, their compressive strength values (50.4–113.4 MPa) do not achieve the minimum requirement for interior use and light duty exterior use. This study delves into the potential of some of our syenogranite samples (I, IIA, IIS, and 10) as gamma radiation shielding materials. We have assessed the mass attenuation coefficient (GMAC), effective atomic number (Zeff), exposure build-up factor (EBF), and energy absorption build-up factor (EABF) for each of these samples. The GMAC and Zeff calculations were performed using the Phy-X online software, across a photon energy range of 0.015–15 MeV. Our findings suggest an inverse relationship between photon energy and GMAC, with the highest values observed for the (I) granite sample (∼18). This study shows the promising radiation shielding capacity of our samples. The insights derived from GMAC, Zeff, EBF, and EABF can serve as a guide for the development of effective, naturally sourced radiation shielding materials.

Novel slag/natural rubber composite as flexible material for protecting workers against radiation hazards

This work is an attempt to employ the electric arc furnace (EAF) slag as a by-product material to develop an alternative and environmentally friendly material for gamma-radiation protection applications such as in medical and industrial areas. For this purpose, different concentrations of micro-sized EAF slag (0, 20, 40, 60, 80, 100, 500, and 800 phr) were incorporated as fillers in the natural rubber (NR) matrix to produce the shielding composites. In addition, nano-sized EAF slag particles were prepared by using a high-energy ball milling technique to investigate the effect of particle size on the gamma-radiation shielding properties. The synthesized micro and nano EAF/NR composites were tested as protective materials against gamma-radiation by employing NaI(Tl) scintillation detector and standard radioactive point sources (152Eu, 137Cs, 133Ba, and 60Co). Different shielding parameters such as linear and mass attenuation coefficient, half value layer (HVL), tenth value layer, mean free path, effective atomic number (Zeff), and effective electron density (Neff) were determined to assess the radiation shielding capability of the EAF/NR composites. Furthermore, equivalent atomic number (Zeq) and the exposure buildup factor values for photon energy in the range from 0.015 to 15 MeV were also computed by Geometric Progression method. The experimental results of micro EAF/NR composites showed that at 121.78 keV, EAF0 composite (without EAF slag content) had the lowest μ value of 0.1695 cm−1, while the EAF800 composite (which was loaded with 800 phr of micro EAF slag) had the highest μ value of 0.2939 cm−1 at the same energy, which in turn decreases the HVL from 4.09 to 2.36 cm, respectively. Therefore, increasing the filler weight fractions of EAF slag in the NR matrix, increases the shielding properties of the composites. Moreover, the NR composite reinforced with 800 phr of nano EAF slag has better gamma-radiation shielding efficiency compared to that filled with 800 phr of micro EAF slag. The success of this work was to prepare a flexible, lightweight, low-cost, and lead-free material with better shielding capability.

Utilization of titanium slag in cement grout for gamma radiation shielding: hydration, microstructure, mechanical properties and gamma-ray attenuation performance

The utilization of industrial waste materials for civil engineering structures has received massive attention in recent decades. In this work, the feasibility of using titanium slag to enhance the radiation attenuation properties of cementitious grout at the ages of 3 days and 2 years was studied. Titanium slag (TS) was incorporated as cement grout replacement at 0%, 10% 20% and 30%. Hardened blocks in dimension of 146×146×30mm were fabricated. The porosity of titanium slag was characterized while the phases in the raw materials as well as the fabricated grout were identified. The influence of the titanium slag addition on the grout hydration kinetics, density, compressive strength and porosity of hardened grouts was evaluated. The results show a decrease in density with the increase of titanium slag content as well as a decrease in compressive strength at curing ages of 3, 7 and 28 days. The slag also refined the microstructure, leading to a reduction in average pore diameter. Furthermore, the hardened blocks were subject to gamma-ray sources such as 137Cs (0.662 MeV) and 60Co (1.173 and 1.332 MeV) and the Linear Attenuation Coefficient (LAC) and Half Value Layer (HVL) were evaluated. The results show that increasing curing age improves the shielding performance of the composite. At 30% replacement, the LAC improvement of the samples exposed to Co sources (1.173 and 1.332 MeV) at 3 days was 11.7% and 12.8% respectively while at 2 years, 9.9% and 9.8% improvement was obtained. Reduced HVL is obtained for an increase in replacement percentage and increased curing age. Therefore, this gives an indication of the possibility of developing low-density radiation shielding cement-based materials.

Gamma radiation shielding properties for polymer composites reinforced with bismuth tungstate in different proportions

In this study, inorganic compound (Bi2(WO4)3) was added into the composite to improve the radiation shielding properties of polymer composite. A polymer matrix was prepared by combining unsaturated polyester resin with methyl ethyl ketone peroxide and cobalt octoate (6%), and Bi2(WO4)3 was added to this polymer matrix at different ratios as filling material. In order to investigate the gamma radiation attenuation properties of the obtained polymer composites, mass attenuation coefficients, radiation shielding efficiencies, radiation transmission factors, linear attenuation coefficients, half values layer, tenth values layer, mean free path values, effective atomic numbers and effective electron densities parameters were obtained. Experimental studies were carried out with the help of HPGe detector at 22 different energies emitted from 22Na, 54Mn, 57Co, 60Co, 133Ba, 137Cs, 152Eu and 241Am radioactive sources in the photon energy range of 59.5–1408.0 keV. Each obtained experimental result was compared with the theoretical results. It was observed that the sample encoded with BiWO20 is the best radiation shielding material among all studied composites (except 59.5 keV).

Physical, optical, and ionizing radiation shielding properties for barium–tellurite glass with different oxides: An experimental study

This study used the melt–quench technique to fabricate four glass systems based on barium–tellurite oxide (BaO–TeO2) glass. The BaO–TeO2 glass system was enhanced with metal oxides such as tungsten oxide (WO3), yttrium oxide, strontium oxide, and lithium oxide. We studied the effects of these oxides as additives experimentally, focusing on their physical, optical, and radiation shielding properties. The fabricated glass density was 4.779–5.372 g/cm3 when different oxides were added. The physical properties of WO3 indicated its superiority over other glasses, supported by the optical properties. We calculated the experimental mass attenuation coefficient for each synthesized sample and compared it to the XCOM software results to assess the radiation shielding efficacy. The difference between the two sets of results ranged from 0.1% to 9.7% for all samples. Furthermore, we calculated the linear attenuation coefficient for each sample, allowing us to determine several properties, such as the effective atomic number, half-value layer, radiation-protection efficiency, mean free path, and transmission factor. According to the gamma radiation shielding results, the sample with WO3 had the best shielding properties, making it a good candidate for radiation shielding applications.

Albedo method applied for gamma radiation shielding

Mathematical tools for calculating radiation shielding usually have difficult notations that could only be solved by computational methods. The Albedo’s method applied for calculation of shielding proves to be an excellent substitute in determining the incident beam fractions that are reflected, absorbed and transmitted, avoiding the use of the transport equation and diffusion approximation that are extremely important in nuclear reactor designs and irradiation calculations. Based on the simple following of the radiation current path, the method can be characterized as a graphical and analytical solution. This work explores the Albedo’s Method applied to 4-slab shielding with incidence of gamma radiation to an energy group compared to ANISN, computational method consecrated in the area of ​​nuclear calculations.

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.

Effects of WO3 reinforcement on the properties of poly(lactic acid) composites for radiation shielding

During this work, efforts have been made to improve the gamma radiation shielding capabilities of poly(lactic acid) (PLA) by including tungsten trioxide (WO3). Simulations and theoretical analyses have been performed on the parameters of radiation shielding. They have been tested across a broad spectrum of energies, from 0.015 to 15 MeV. The values suggest that naturally occurring composites with a high WO3 content may be utilized to improve the gamma shielding properties of PLA/TO (TO ≡ tungsten trioxide) composites for protection applications; such composites can be employed as a covering for the solar cells to prevent damage. The composite samples made of PLA and WO3 were analyzed for a range of photon energies. The results showed rise of considerable proportions in the values of linear attenuation coefficient (μ) and mass attenuation coefficient μm at 0.08 MeV, due to the k-absorption edge of tungsten. In addition, μ and μm for all samples at selected photon energy are increase as the WO3 content increases and decreases with increasing photon energy. The half value layer (T0.5) decreased as the WO3 content increased, with the highest values seen at 15 MeV and the lowest values at 15 keV. The behavior of the mean free path (λ) was identical to T0.5 with regards to the amount of WO3 and the energy of the photon. The T0.5 values of the PLA-W7 were lower than other polymer materials, making it a more efficient material for filtering radiation. The Zeff values of the PLA/TO composite samples were the highest at 0.08 MeV and showed a trend of increasing with energy to 15 MeV. The exposure buildup factor (EBF) and energy absorption buildup factor EABF values were computed using the G-P fitting method and showed that the lowest values were presented for the PLA-W7. An increase in WO3 reinforcement in the PLA sample contributed to the decreased EBF-EABF values. The shielding parameters were determined through calculations done using both Phy-X/PSD and FLUKA, and the results from both are in good agreement with each other. The results of the study demonstrate the effectiveness of PLA-W7 as a material for radiation shielding.

Exploring the potential of BBNCo glasses: Physical, optical, and radiation shielding analysis

The ability of new Co3O4-doped bismuth borate glasses to block radiation was investigated in this study. The glasses had compositions of 55B2O3–35Bi2O3-(10-x)Na2O-xCo3O4(where x = 0, 0.75, 1.5, 2.25, and 3 wt%). The optical properties of the prepared bismuth borate glass doped with Co3O4ions have been studied. The UV demonstrates that as the Co3O4substitution ratio was increased, the samples' absorbance increased noticeably. The optical band gap energies decreased with increasing Co3O4concentration in the case of direct transition and in-direct transition. In addition to the Urbach energy EUvalues of the prepared glass samples was decreased as the concentration of Co3O4ions increased. Using the Phy-X/PSD program, an investigation was conducted on the gamma-radiation shielding capabilities of new BBNCo glasses. The GLAC (linear attenuation coefficient) readings for BBNCo increased from a previous value of 0.7575 cm2/g to a new value of 0.8589 cm2/g when the energy was 356 keV. In addition, computations have been made to determine the removal cross-sections of fast neutrons and the neutron half value layer (NHVL). The addition of Co3O4seems to increase the efficiency of bismuth borate glass in deflecting gamma and neutron radiation, as shown by the findings.

Development of novel nanocomposite radiation shielding blocks as gamma rays barrier

This work is focused on developing new types of radiation shielding materials using the stereolithography 3D printing technique, which enables the creation of gamma radiation shielding materials in two different sizes. Six different square nanocomposite radiation shielding blocks with an additive ratio of 0.5 wt% were prepared using aluminum nano-powder (nano-Al), copper nano-powder (nano-Cu), calcium nano-powder (nano-Ca), silicon carbide nano-powder (nano-SiC), graphene nano-platelets (GnPs), and eggshell nano-powder (nano-Es). Then, blocks prepared in two different thicknesses (5 mm and 10 mm) were analyzed with Mössbauer spectroscopy utilizing the Co-57 source for γ-ray shielding. The radiation attenuation properties of the developed materials for γ-rays were evaluated. The nano-Cu doped composite block was determined to have the greatest performance in gamma-ray shielding among the specimens mentioned in this investigation. The findings indicate that it is possible to reduce the shield thickness by about 19% compared to neat resin for γ-rays using the determined nano-powders. These experimental findings suggest that the nano-Cu, nano-Ca, and nano-Al reinforced resin nanocomposites are potentially valuable new radiation shielding materials.

Fabricated CeO2/ZrO2 nanocomposite to improve thermal resistance, mechanical characteristics, microstructure and gamma radiation shielding of OPC composite cement pastes

The main objective of this study is to utilize the synthesized CeO2/ZrO2 nanocomposite (CZ-NPs) in different concentrations (0.25, 0.5, 0.75, 1, and 1.5%) for improving the mechanical characteristics, microstructural properties, thermal resistance, and gamma radiation shielding of hardened ordinary Portland cement (OPC) composites. CZ-NPs were fabricated via the precipitation method. Various analytical techniques in addition to X-ray diffraction (XRD), Fourier-transform infrared spectroscopy (FTIR), Brunauer-Emmett-Teller (BET), and Scanning electron microscopy (SEM) analysis were used to investigate the fabricated CZ-NPs and hardened composite pastes. The physico-mechanical features, thermal stability, and radiation resistivity of different OPC-CZ-NPs composite pastes were investigated. The results showed that the composite CZNP1 containing 0.25% CZ-NPs is the best choice for mechanical performance and has the highest gamma-ray shielding efficiency, as it can attenuate 50–90% respectively of gamma rays with energy 660Kev at distances 4 cm and 13 cm, respectively. Also, the composite accommodating of 0.5% CZ-NPs motivates the thermal resistance at 400 °C.

Design of a radiation shield for an HPGe detector for a reactor-based PGNAA facility

The purpose of this study was to design an appropriate neutron and gamma radiation shield for the HPGe detector to be used in the PGNAA facility in the Isfahan MNSR reactor. In order to improve the parameters related to shield design, including the thickness of the material used in the design, the different type of materials considered and the relative reduction of neutron and gamma dose, shields were simulated by MCNPX code and the results of calculations were compared. It was concluded that the optimal thickness of lead as gamma absorber is 4 cm and the optimal thickness of borated polyethylene as neutron absorber is 10 cm. In this case, the neutron and gamma dose rate at the location of the detector inside the shield was equal to 5.11 μSv.h−1 and 7.5 μSv.h−1, respectively. Also, according to the investigations conducted for different materials as neutron absorbers, it was found that the best absorber for neutrons is borated polyethylene, which absorbs more neutrons and produces less secondary gamma than other materials and the best absorber for gamma is bismuth. The shielding has brought the relative percentage of neutron and gamma dose reduction about 90%.

Investigation of the effect of cement type on nuclear shield performance of heavy concrete

This article studies the radiation shielding performance of three different concrete-added samples. First, the mass attenuation coefficient values of three distinct concrete materials were computed using the MATLAB code at energies ranging from 0.015 to 15 MeV, and the findings obtained were verified by theoretical WinXcom results. Half-Value Layer (HVL), Tenth-Value Layer (TVL), Mean Free Path (MFP), Effective Atomic Number (Zeff), Exposure Build-Up Factor (EBF), Energy Absorption Build-Up Factor (EABF) are the variables computed to assess available concrete samples in terms of radiation attenuation characteristics. The results demonstrated that the concrete sample with CAC possesses better gamma-radiation shielding characteristics. Concrete samples’ removal cross-sections (∑R) were obtained to analyze their ability to stop fast neutrons. Iron, aluminum, and lead-added concretes are better in terms of their density than the other types of concretes. Given the relationship between the density of the material and neutron shielding propensities, iron, aluminum, and lead-added materials are expected to better serve for shielding purposes. SRIM Monte Carlo code, as it is standard, was utilized to compute the mass stopping power (MSP) and projection range (PR) values. The concrete sample with CAC has shown better shielding properties compared to the others in the sample of the concretes having the lowest HVL, TVL, and MFP and highest ∑R. For proton and alpha particles, the concrete sample with CAC sample shows the best radiation shielding and protection properties. The concretes selected for the study are useful for shielding purposes compared to many other materials used in industry. In addition, radiation shielding concrete is cheaper than other shielding materials. It is easier to mold into different geometric shapes and is suitable for shielding from protons or alpha particles. Among the different concretes investigated in this study, it can be used for structures such as radiation oncology and nuclear medicine clinics in hospitals. Additionally, it can act as a biological shield for nuclear power plants, particle accelerators, and radionuclide laboratories, as well as other sources of radiation.