Total Mass Attenuation Coefficients Research Articles

The possibility of using the DKS-AT1123 dosimeter for radiation monitoring of medical electron accelerators with the energy of more than 10 MeV

In the Russian Federation, there is a constant increase in the number of radiation medical installations with electron accelerators. Over the past 4 years, their number has increased 2.5 times. These installations contain pulsed electron accelerators that generate pulsed bremsstrahlung radiation with a maximum energy from 6 to 21 MeV. Currently, there are no devices designed for dosimetry of the pulsed photon radiation with energy of more than 10 MeV in the state register of measuring instruments of the Russian Federation. The most widely used radiation monitoring device for pulsed electron accelerators is the DKS-AT1123 X-ray and gamma radiation dosimeter designed for dosimetry of pulsed bremsstrahlung radiation with an energy of up to 10 MeV. The purpose of this work is to evaluate the possibility of using this device for dosimetry of pulsed bremsstrahlung radiation with a maximum energy of up to 20 MeV. The authors calculated the energy spectra of bremsstrahlung radiation for a point source with a maximum energy of 20 MeV behind flat concrete screens with a thickness of 1 m, 2 m and 3 m by the Monte Carlo method using the GEANT4 calculation program. The energy dependence of the registration efficiency of the DKS-AT1123 dosimeter was extrapolated to the energy range of 10–50 MeV in the kerma-approximation without taking into account the energy transfer by secondary electrons. It was assumed that it corresponds to the energy dependence of the total mass attenuation coefficient for the absorbed energy of gamma quanta in water. Using conversion coefficients for converting the fluence of monoenergetic photons into the effective dose rate at an anterior-posterior radiation incidence on the human body, real dose rates were calculated, and using the energy dependence of the dosimeter readings, the predicted results of measuring the unit dose rate with the DKS-AT1123 dosimeter behind a concrete protection with a thickness of 1, 2 and 3 m were obtained. It is shown that the maximum expected underestimation of the measurement results will not exceed 40% and practically does not depend on the thickness of the concrete shield in the thickness range from 1 to 3 m. To account for this underestimation, it is necessary to use the value of additional measurement error due to the energy dependence of the sensitivity of this device for the photon radiation energy of more than 10 MeV, equal to 70%. This makes it possible to use the measurement results obtained using this dosimeter to adequately characterize the state of radiation safety during operation of pulsed electron accelerators with a maximum energy of up to 20 MeV. It is possible to use a correction factor to the measurement results equal to 1.63 ±0.04 to compensate for this underestimation. The proposed approach can be used to create a methodology for using this dosimeter for radiation monitoring of medical electron accelerators with the energy of up to 20 MeV, if there are correction factors for radiation protection configurations and radiation energies encountered in practice.

Developed concrete reinforced by waste of lead fishing weights and steel nails as gamma rays and neutrons shields for nuclear power reactors

Wastes of steel nails WSN and wastes of lead fishing weights WLFW were successfully recycled in concrete to enhance its ability to attenuate neutrons and gamma rays. A series of studies on the effect of partially replacing coarse aggregates of concrete with different percentages (5, 10, and 20%) of WSN or WLFW on the density, slump, compressive strength, and attenuation of neutrons and gamma rays were implemented. The compressive strength of WSN-reinforced concrete ranged between 36.2 and 45.2 MPa, while that of WLFW-reinforced concrete ranged between 28.6 and 45.1 MPa. The ability of WSN- or WLFW-reinforced concrete to attenuate neutrons and gamma rays was tested by using two types of neutron energies; slow and total slow neutrons; and three gamma rays energies 661.64, 1173.23, and 1332.51 keV. Slow and total slow neutrons macroscopic cross-section, linear and mass attenuation coefficients of gamma rays, and half value layer for both types of nuclear radiation were obtained. The results showed that there is a significant improvement in the slow and total slow neutron attenuation efficiency of the considered concrete when reinforced by WSN or WLFW. On the other hand, due to the increase in the atomic number and density of both Fe present in WSN and Pb in WLFW, the attenuation ability of the studied gamma ray energies increased. The WLFW-reinforced concrete showed better value in the gamma rays and neutrons attenuation parameters compared to the WSN-reinforced concrete. Hence, the considered concretes reinforced by WSN or WLFW are a distinguished choice as radiation shields in nuclear power reactors, in addition to their adaptability to high concentrations of waste.

Radiation shielding for inertial electrostatic confinement fusion system utilizing concrete and water

In this investigation, the impact of water thickness and various types of concrete materials, each characterized by different densities and elemental compositions, is examined for their role in shielding a radiation generator based on Inertial Electrostatic Confinement Fusion (IECF) device. The study focuses on assessing several vital parameters, including the effective removal cross-section (∑Rt) for fast neutrons, total mass attenuation coefficients (μmt), linear attenuation coefficients (μl), half-value layer (HVL), and mean free path (MFP) for X-rays across different concrete types. Different water thicknesses around the IECF chamber, ranging from 0.5 to 10.0 cm, are investigated, and five concrete types are evaluated: Ilmenite-magnetite Concrete (IMC), Ordinary Concrete-1 (OC1), Barite-containing Concrete (BC), Ordinary Concrete-2 (OC2), and Serpentine-containing Concrete (SC). The results indicate that, among these materials, SC requires the least thickness to attenuate 2.45 MeV generated from IECF to 1/100th of its initial intensity across varying water thicknesses (6.5 cm in case of 5 cm water thickness). The values of μmt, HVL, and MFP are also calculated for different water thicknesses and X-ray energies (ranging from 0.2 to 3.0 MeV). These calculations highlight BC as the material requiring the least thickness to attenuate X-rays to 1/100th of their initial intensity. Moreover, neutron dose rate measurements are conducted on a commercial IECF system shielded with 50 cm of water and operated at a neutron intensity of 105 ns−1, which was ∼12 nSv/h on average, approximately 0.002% of the initial intensity. This underscores the efficacy of water shielding in attenuating the outcomes of IECF.

Long-term stability of hydroxyapatite bone phantoms for the calibration of in vivo x-ray fluorescence spectrometry-based systems of bone lead and strontium quantification

Hydroxyapatite (HAp) phantoms have been proposed as an alternative to plaster of Paris (poP) phantoms for the calibration of x-ray fluorescence-based systems for the in vivo quantification of bone lead and strontium which employ a coherent normalization procedure. The chemical composition of the material becomes critical in the calculation, or omission, of the coherent correction factor (CCF) required in this normalization procedure. This study evaluated the long-term chemical stability of HAp phantoms. Phantoms were prepared and allowed to age for a two week period and over a seven year period in ambient conditions. The chemical composition of the phantoms was then assessed by powder x-ray diffraction. Two week old phantoms were found to be composed of HAp with only a small amount of contamination from CaHPO4·2H2O. Seven year old phantoms were found to have converted nearly completely to a carbonate-bearing apatite in the form of Ca10(PO4)6(CO3)0.75(OH)0.5 indicating that the HAp phantom material likely reacts with carbon dioxide in air over time forming a carbonate-bearing apatite. The influence of this chemical conversion was assessed at the level of relevant cross-sections. Calibration under the assumption that the material is HAp when in fact it is a carbonate-bearing apatite would result in not more than a 0.2%–2% bias in the total mass attenuation coefficient within the photon energy range of 0–100 keV. Differential scattering cross-section for coherent scattering was found to differ between HAp and carbonate-bearing apatite by 0.9%–2% for both a 35.5 keV and 88.0 keV γ-ray. This variation in the differential scattering cross-section for coherent scattering may introduce a ca. 2% bias in the CCF used within the coherent normalization-based calibration procedure. Using HAp phantoms as calibrators thus requires acknowledgement of this conversion in chemical form and possible introduction of uncertainty into the calibration procedure.

Assessment of Five Concrete Types as Candidate Shielding Materials for a Compact Radiation Source Based on the IECF

A radiation source based on the inertial electrostatic confinement fusion (IECF) system is being developed for multidisciplinary research applications. The radiation outputs from the IECF system are 2.45 MeV fast neutrons and the associated co-generated X-rays with an energy less than 3 MeV. A radiation shielding study has been performed on five types of concrete to define the most efficient material for the shielding design of the system. The proposed materials were ilmenite-magnetite concrete (IMC), ordinary concrete-1 (OC-1), barite-containing concrete (BC), ordinary concrete-2 (OC-2), and serpentine-containing concrete (SC). A numerical model was applied to determine the effective removal cross-section coefficients (∑Rt) for the fast neutrons and the total mass attenuation coefficients (µm), the half-value layer (HVL), the mean free path (MFP), the effective atomic number (Zeff), and effective electron density (Neff) for photons inside the materials. The model considered the radiation source energy and the material properties of the concrete types. The results revealed that the serpentine-containing concrete exhibited the highest ∑Rt with 12 cm of concrete thickness needed to attenuate an incident neutron flux to 1/100 of its initial value. In addition, the BC shows the highest µm with a 38 cm concrete thickness needed to attenuate the 3 MeV energy X-ray flux to 1/100 of its initial value. This study suggests that a 40 cm thickness of SC or BC adequately shields the radiation generated from an IECF system with a maximum particle production rate of up to 1 × 107 n/s.

A new approach to calculating the ratio of the compton to total, mass attenuation coefficient

X-rays and gamma rays are high-energy radiation that ionize atoms of matter as they pass through them. These rays interact with matter in such a way that secondary rays with energy equal to or lower than the incident photon are produced. The buildup factor is used to estimate the intensity of these secondary rays. One of the methods to calculate the buildup factor is the Geometric Progression formula. To use this formula, the equivalent atomic number must be determined. The ratio of the Compton mass attenuation coefficient to the total mass attenuation coefficient (R) is used to calculate the equivalent atomic number for each compound, element, or mixture. In this research, an equation was introduced to calculate the ratio of the Compton mass attenuation coefficient to the total mass attenuation coefficient (R) for X and gamma rays. This equation was the first of its kind and had good accuracy. It was a two-variable function of energy and atomic number that covered the energy range of 0.015–15 MeV and the range of 4–92 for atomic number. The highest SSE and RMSE fitting errors for the energy component for all calculated elements were equal to 0.000921 and 0.0076, respectively, and the same calculation errors for the atomic number component parameters were equal to 9.5561 and 0.7092, respectively. The maximum relative error of the calculated value with the actual value of water as a composition was less than 6%, and the amount of the same error for lead as an element was less than 9%. This equation can be used in other nuclear codes, as the functions used in it are very straightforward. It reduces the speed of calculations and the volume of calculations for calculating the secondary radiation dose.

Photon attenuations of trichloride of gadolinium by Geant4 and XCom: a comparative study

The total mass attenuation coefficient of trichloride gadolinium has been investigated utilizing Geant4 compared with WinXCom simulation data at energies; 1–10 MeV. The simulated results are used to calculate the total cross section, atomic and electronic cross sections. The aim of this work is to characterize new materials and their interactions with radiation in order to be able for using in shielding or detection. The GdCl3 is applied for super–Kamiokande detector of neutrons. From our comparison study, it is found that the mass attenuation coefficient of GdCl3 with those of two detectors of X-ray (Si and CdTe) is very closed. Also, the total mass transfer attenuation coefficient increases as high energies increases at pairs effect production. From our comparison study, we can predict that GdCl3 can be a good detector of X-ray and concluded that this material has shown good interaction with electromagnetic radiations at high energies.

A study for gamma-ray attenuation performances of barite filled polymer composites

In this study, radiation protection efficiency (RPE) for the coded as UP-Ba0, UP-Ba25, UP-Ba50, UP-Ba75 and UP-Ba100 at different sample thicknesses, total mass attenuation coefficient (μ/ρ), linear attenuation coefficients (μ), half value layers (HVL), tenth value layers (TVL), mean free paths (MFP), effective atomic numbers (Zeff) and effective electron densities (NE) were determined at various gamma energies between 59.5 and 1408.0 keV. With the help of the geometric progression (G-P) fitting method, the energy absorption build-up factor (EABF) and exposure build-up factor (EBF) values were calculated in the energy range from 0.015 MeV to 15 MeV for the produced composites. HPGe detector and eight radioactive sources (241Am, 152Eu, 137Cs, 133Ba, 60Co, 57Co, 54Mn and 22Na) were utilized in the experiment. Experimental results were compared with theoretical calculations and it has been observed that there is a good agreement between theoretical and experimental results. It was observed that RPE, μ/ρ, μ, Zeff and NE parameters increased with increasing barite amount and decreased with increasing energy, while the opposite situation was observed in HVL, TVL and MFP parameters. EABF and EBF values increase with increasing penetration depth. As a result, UP-Ba100 is a good radiation absorber according to the other studied barite filled polymer composites.

Shielding Properties of Glass Samples Containing Li2O, K2O, Na2O, PbO and B2O3 by Geant4, XCOM and Experimental Data

Abstract: In the present work, glass samples containing 10Li2O, 10K2O, 20Na2O. xPbO, (60-x)B2O3 (where x = 0-60) were prepared by the melt quenching method. The shielding parameters of the prepared samples were measured experimentally and calculated theoretically. The measurements have been performed at energies of 356.5, 662, 1173 and 1333 keV to obtain the total mass attenuation coefficient (μm), using a gamma spectrometer containing a shielded NaI (TI) detector. The results of mass attenuation coefficient (μm), half-value layers (HVLs), mean free path (MFP), radiation protection efficiency (RPE), atomic and electronic electron cross-sections (σa and σe), effective atomic number (Zeff) and effective electron number (Neff) were calculated at energies from 1 keV to 100 GeV using the Monte Carlo simulation code Geant4 and XCOM. The calculated results were compared with each other and with the experimental values. Good agreement has been observed. Keywords: Shielding properties, Glass, Photon mass attenuation coefficient, Atomic and electronic cross-sections, Effective atomic and electron numbers, XCOM, Geant4.

The influence of antimony additive on structural, mechanical, and nuclear radiation shielding parameters of rapidly quenched Pb-Sn binary alloy

This study aims to investigate the ability of as-prepared PbSn3Sb x (x = 0.0, 0.5, 1.5, and 2.5 wt.%) alloys for shielding against gamma-rays, the relationships between the mechanical properties and gamma-ray shielding parameters for four types of alloys have been studied and analyzed. Gamma-ray interaction with the as-prepared alloys has been discussed mainly in terms of total mass attenuation coefficient (µm ), linear attenuation coefficient (µ), half value layer (HVL), and mean free path (Xm ). From X-ray analysis, a crystalline intermetallic compound phase of SnSb is detected. The formation of the intermetallic compound phase and refinement of evaluated particle size causes a pronounced enhancement in internal friction, dynamic elastic modulus, and microhardness indentation. Attenuation shielding parameters of as-prepared alloys with free antimony are measured and calculated. It is noticed that the obtained results showed desirable values of shielding parameters. Mechanical properties and Vickers microhardness, Hv are critically evaluated and sensitive to changes in temperature and Sb antimony additions. Thus, the study indicates that 2.5 wt.% Sb alloy sample has a distinguished protection ability to attenuate the gamma-ray radiation. It can be concluded that the Sb additive for this type of binary alloys at content 2.5 wt.% can help to increase the shielding gamma-ray performance.

Effect of Gd2O3 on radiation shielding, physical and optical properties of sodium borosilicate glass system

The radiation shielding, physical and optical properties were studied with the glass formula in the composition of xGd2O3: 40Na2O: 5SiO2: (55-x)B2O3, where x = 0, 2.5, 5, 7.5, 10, 12.5, 15 mol%, respectively. The gadolinium sodium borosilicate glasses were melted entirely by melt quenching technique at 1200oc. Gamma-ray shielding property was studied with Compton scattering technique from 0.224 MeV to 0.662 MeV gamma-ray energy range. The density and molar volume of the glass samples were increased with higher Gd2O3 concentration. The glass samples show high transparency which can be observed from transmission spectra. The radiation shielding properties are better with increased Gd2O3 concentrations. The results of total mass attenuation coefficient, the effective atomic number and the effective electron density showed the theoretical and experimental values are in good agreement and showed similar trends. The best shielding radiation properties were obtained at 15% of Gd2O3 concentration. At 15% of Gd2O3 concentration shows the same attenuation behavior with 0.64 mm of lead. The half value layer (0.662 MeV) at 15% of Gd2O3 concentration is better than the commercial window, serpentine and ordinary concrete.

Proton, Alpha, and Gamma Rays Interactions of CsI(Na) Scintillator Using the Theoretically Computational Program

The charged particles and photon interaction parameters of CsI(Na) scintillator were calculated by the theoretically computational program. The Phy-X programs were used to calculation gamma rays interaction at an energy range of 1 keV to 100 GeV. The stopping and range of ions in matter program were used for measurement the proton and alpha particles. The result found that the total mass attenuation coefficient decreased with the increasing of the high photon energy range and found the discontinuous energy due to the absorption edge of Cs and I. The effective atomic number and the effective electron density found the same trend with the total mass attenuation coefficient. The data of mass stopping powers for proton and alpha particle increase to the maximum value at kinetic energy 0.09, 0.6 MeV, respectively. The results showed that the energy at 5 keV, the ions entering the scintillator penetrate less deeply than the energy of the ions at 10 keV.

Characterization of new materials for protection against ionizing radiation in radiology

The use and manipulation of radiation sources and equipment generating ionizing radiation in various applications of science, industry and medicine have increased significantly in recent years. This extensive use of these sources can lead to a longer time radiation exposure, eventually causing possible health effects for medical personnel. Therefore, the use of shielding materials for protection is paramount, reducing the risk of radiation exposure. The objective of this work was to manufacture three new shielding materials against ionizing radiation and compare them with other protective materials that already exist commercially: concrete galena, standard concrete and those manufactured by the Shieldwerx Company (a division of Bladewerx LLC). These new materials were called Sice, Sicer and Sire. Parameters such as the total mass attenuation coefficient , half-value layer and effective atomic number () for energies of 1 keV to 1 GeV were simulated using the PENELOPE Code (2014 version) and the Auto-Zeff software (1.7 version). To simulate these new materials, it was necessary to previously determine their density, chemical composition and percentage by the atomic weight of each element which was analyzed by scanning electron microscopy (SEM). It was concluded that Sice and Sicer materials have greater absorption against ionizing radiation compared to the other materials of the Shieldwerx company and Standard concrete. For energy values of 25 and 84 keV, it was found that Sicer absorbs 25% more radiation than the Shieldwerx Company’s best material (field-castable head-resistant shielding) and Standard concrete. A slightly lower percentage was found for the Sice material with the same energy values. Therefore, these new materials can be potential shielding materials in mammography (24–30 keV) and conventional radiography (40–150 keV).

The Calculation of Photon Interactions in Gd2YAl2Ga3O12 compound by the Phy-X Software Compared with WinXCom Program

The photon interactions in Gd2YAl2Ga3O12 compound were studied in this work. The interaction parameters, such as total mass attenuation coefficient (μm), effective atomic number (Zeff), effective electron density (Neff), mean free path (MFP), half-value layer (HVL) and tenth-value layer (TVL) were computed as a function of energy between 1 keV-100 GeV, while the atomic equivalent (Zeq) and the buildup factors were calculated with 0.015 MeV-15 MeV energy range using Phy-X software. Moreover, all obtained values were compared with calculated ones using WinXCom program over the same energy range. Both calculation processes represent the μm reduction by following the energy increment. The Zeff and Neff show a similar trend depending on the energy range and have some discrete values at low energy. These discontinuous values correspond to the absorption edges which were also found in the μm calculation at same energy. The trends of MFP, HVL, and TVL are similar in that their highest values appear in the middle energy range with main interaction being Compton scattering. The buildup factors represent the multiple scattering of photons in materials and air with the main peaks in middle energy range.

A study on radiation shielding potentials of green and red clayey soils in Turkey reinforced with marble dust and waste tire

The increasing radiation applications in our daily life makes it essential to protect ourselves from the harms of radiation by using alternative, cheap and natural materials. The present study aimed to analyze the radiation shielding abilities of green and red clayey soils from Oltu/Erzurum in Turkey, reinforced with waste tires and marble dust. For the purpose to investigate the shielding features of the samples, radiation attenuation parameters were determined by using EpiXS software, which can calculate partial or total cross-sections, partial or total mass attenuation coefficients, electron densities, effective atomic numbers, and buildup factors for energy absorption and exposure between 1keV and 1GeV. We compared the obtained mass attenuation coefficients and total atomic cross-section values of the samples with those of a widely used shielding material, ordinary concrete, to make a meaningful evaluation about the shielding potentials of the samples. To validate obtained values by EpiXS, we also calculated the mass attenuation coefficients of the samples by XCOM code, and compatible results were obtained. Among all the studied clayey soil samples, green clay reinforced with marble dust and waste tire has the highest shielding capability. It can also be mentioned that reinforcement with marble dust and waste tire improves the shielding ability of the clayey soils.

INFLUENCE OF DIFFERENT NATURAL FIBER-BASED COMPOSITES USE AS A FILLER FOR CRACKING IN NUCLEAR REACTOR BIOLOGICAL SHIELDS

The biological shield is very important in the nuclear power reactor, so it is necessary to study the effect of crack on it. This work aims to study fiber composites as a radiation shielding for various purposes associated with nuclear installations, as properly as a repairing mixture for developing cracks in the biological shields of a nuclear power reactor. In this study, the MCNP-5 model of a pressurized water reactor (PWR) was used. Four different fiber composites of natural fiber (FP) (ρ= (1.373 g/ m3), fiber with lead (FPPb) (ρ=2.756 g /cm3), cement fiber (CF) (ρ=2.095 g/ m3) and cement-fiber-magnetite (CFM) (ρ =2.858 g /cm3) were used as the filler materials for cracking in the biological shield. The total mass attenuation coefficients µm for the studied composites have been calculated by the Win X-Com program of gamma-ray at an energy range from 0.15 – 20 MeV. Neutron and gamma-ray fluxes and dose rates were calculated through different distances of the reactor shields for different fiber composites. Total dose rates through biological shield without cracking and after cracking have been calculated. Radiation Protection Efficiency (RPE %) for all samples at the outer surface of the biological shield was calculated. The results showed that by using fiber composites as a filler for cracking, the values of flux and dose rate for neutrons and gamma-rays have decreased. Also, FPPb is the best material for gamma-ray because it contains lead, which has a high atomic number. While CFM has the best material shield for neutron and gamma-ray in addition to magnetite and boron carbide the properties of CFM composite were improved. Max. value of RPE % for FPPb was (43.1%), then CFM (37.2%), Fp (30.2%) and CF (20.2%). From these results, it can be concluded that plastic fiber composites were the best materials that can be used as a filler in the biological shield of nuclear power reactors.

High density of tungsten gadolinium borate glasses for radiation shielding material: Effect of WO3 concentration

The aim of this work is to study the radiation shielding properties in the glass systems xWO3: (70-x)Gd2O3: 30B2O3 (WGB) where x = 40, 45, 50, 55, and 60 mol%. These glasses were prepared by the melt quenching techniques. The Compton scattering technique was used to measure the experimental values of the total mass attenuation coefficient (MAC), the effective atomic number (Zeff), and the electron density (Neff) at 0.223 MeV–0.662 MeV. The measured MAC, Zeff, and Neff results were compared with the simulated and theoretical values obtained from Geant4 code and WinXcom software respectively. The comparison between the experimental, simulation and theoretical approaches for the determination of the several radiation shielding factors reflects the good detection system setup of Compton scattering experiment. The density is decreased with the increasing of WO3 concentrations from 6.1868 g/cm3 to 5.2669 g/cm3 and molar volume result demonstrated non-bridging oxygen and network former in the glass structure. The MAC, Zeff, and Neff results have similar trends, which increase by adding WO3 contents, and decrease with increasing of the energies. The HVL result showed that increased at the higher WO3 concentrations. Moreover, the HVL at 40 mol % of the concentration showed the highest density, and demonstrated better efficiency when compared with some standard shielding materials.

Neutron/gamma radiation shielding characteristics and physical properties of (97.3−x)Pb−xCd–2.7Ag alloys for nuclear radiation applications

In this work, the shielding performance of (97.3–x)Pb–xCd–2.7Ag (x = 10, 18, and 30) ternary alloys against neutrons and gamma rays has been investigated. The microstructure, thermal and mechanical properties of the ternary alloys were examined. The total mass attenuation coefficients, for prepared alloys were determined at 662, 1173, and 1332 keV photon energies using NaI (Tl) scintillation detector. The theoretical values of were calculated using WinXCom program depending on the mixture rule. The estimated values were compared with the measured values for all investigated alloys. Atomic cross-section, σ a, electronic cross-section, σ e, effective atomic number, Z eff , effective electron number, N eff , and GP fitting parameters (b, c, a, Xk, and d) were determined. The exposure buildup factor, EBF, have been also calculated. Fast neutron attenuation for the prepared samples have been investigated via the macroscopic effective removal cross-section () calculation. Also, thermal neutron attenuation has been evaluated via neutron scattering calculator. The results show that the alloys containing 10 and 30% Cd compromise between superior tensile strength and Young modulus, while the pasty range, heat of fusion and ductility decreased with increasing Cd content. Moreover, the prepared ternary alloys have a high attenuation ability for gamma rays as the standard Pb. The increase of Cd ratio also significantly enhances the thermal neutron attenuation by amazing way along with the increase in the attenuation rate of fast neutrons.

Gamma radiation shielding performance of CuxAg(1-x)-alloys: Experimental, theoretical and simulation results

Different types of photon shielding parameters such as total mass attenuation coefficient (μ/ρ), linear attenuation coefficients (μ), half value layers (HVL), mean free paths (MFP), effective atomic numbers (ZEff), energy absorption build-up factors (EABF), exposure build-up factors (EBF) and kerma relative to air were investigated for the fabricated Cu–Ag based alloys. The considered parameters were measured through gamma spectrometer equipped with HPGe detector in order to obtain the experimental attenuation coefficients and other related parameters at various photon energy in the energy range 59.5–1332.5 keV. The measured μ/ρ values were confirmed with WinXCOM database results. FLUKA and GEANT4 simulation codes were used to examine the compatibility of the experimental and WinXCOM database results with these simulation codes. The exposure buildup factors of the alloy samples were estimated with help of Geometric Progression fitting formula over photon energy 0.015–15 MeV up to 40 mfp penetration depth. The results revealed that the exhibited effectiveness of Cu0.2Ag0.8 alloys against high energetic photon radiations had a good performance than that of alternative absorbers such conventional concretes, glasses and some alloys. The results of the present survey can be quite useful for possible applications of such materials, especially in nuclear laboratory and reactor core design for preference of effective photon shielding materials.

Ion bombardment technique induced alterations in the surface chemical composition and shielding parameters of polymeric material

At present, ion beam technology is one powerful technique to improve the physico-chemical properties of polymeric material. In this work, ultra high molecular weight polyethylene samples were bombarded with 1 MeV He-ion beam at fluences ranging from 1×1013 to 5×1014 ions/cm2. The changes in the chemical composition of the treated layer caused by the ion beam were investigated. The hydrogen escapes due to the ion beam irradiations were measured by the nuclear reaction analysis (NRA) technique using the 1H(15N, αγ)12C reaction. The oxygen uptake has been studied by the Rutherford backscattering spectrometry (RBS) and Fourier transform infrared spectrometer (FTIR) techniques. The partial and total mass attenuation coefficients of photons at energies 81, 356, 511 and 1332 keV were estimated using the XCOM program (version 3.1). The obtained results showed that the hydrogen release was increasing with an increase in ion fluence. Moreover, important oxygen uptake of bombarded samples was also observed. These effects led to significant improvements in the surface characteristics of the treated materials such as changes in the shielding parameters. The changes in the mass attenuation coefficients (μ/ρ) were calculated relying on the incoming photon energies and the atomic concentrations of the elements that are present in the target material. The total atomic cross-sections and effective atomic numbers were calculated by using the data of the total mass attenuation coefficients. A decrease in both the total atomic cross-section and effective atomic numbers was noticed with an increase in the ion beam fluence.