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In this study, the fast neutron and gamma-ray absorption capacities of the new glasses have been investigated, which are obtained by doping CoO,CdWO4,Bi2O3, Cr2O3, ZnO, LiF,B2O3 and PbO compounds to SiO2 based glasses. GEANT4 and FLUKA Monte Carlo simulation codes have been used in the planning of the samples. The glasses were produced using a well-known melt-quenching technique. The effective neutron removal cross-sections, mean free paths, half-value layer, and transmission numbers of the fabricated glasses have been calculated through both GEANT4 and FLUKA Monte Carlo simulation codes. Experimental neutron absorbed dose measurements have been carried out. It was found that GS4 glass has the best neutron protection capacity among the produced glasses. In addition to neutron shielding properties, the gamma-ray attenuation capacities, were calculated using newly developed Phy-X/PSD software. The gamma-ray shielding properties of GS1 and GS2 are found to be equivalent to Pb-based glass.

Using the sol–gel process, a series of Sr2-xZnSi2O7:xGd (0.01 ≤ x≤0.11) samples was fabricated. Their crystal characteristics, surface morphologies, and spectral characteristics were analyzed using X-ray powder diffraction (XRD), scanning electron microscopy (SEM), and photoluminescence (PL) spectroscopy. Upon 272-nm excitation, a prominent emission band appeared at 313 nm (6P7/2 → 8S7/2) in all the samples. The electron paramagnetic resonance (EPR) and PL analyses confirmed the presence of Gd3+ in the Sr2ZnSi2O7 structure, where the Gd3+ ions occupied distorted Sr2+ sites in the host lattice. Additionally, three photon interaction parameters, namely the mass attenuation coefficient, effective atomic number, and effective electron density, were calculated within the 1 keV–100 GeV photon energy range for all the samples. These photon interaction parameters varied widely over the studied energy range. Particularly, the shielding effectiveness increased with increasing gadolinium concentration. These interaction parameters will be very useful for shielding applications against gamma-rays.

This work aimed to investigate the mechanical features and radiation shielding properties of (100–2x)TeO2+xAg2O + xWO3 (where x = 7.5, 15, 22.5, and 30 mol%) glass system. Based on Makishima–Mackenzie's theory, the basic mechanical features such as packing density, Poisson's ratio, elastic moduli, hardness, and fractal bond connectivity were studied. The shielding investigations included the radiations of gamma, electron, neutron, and heavy charged particles. Geant4 toolkit and newly developed Phy-X program were employed to evaluate mass attenuation coefficient (μ/ρ), transmission factors (HVL and MFP), exposure buildup factor (EBF) and effective atomic number (Zeff) of gamma and charged particles for the studied glasses. Neutron shielding investigation was examined by determining neutron removal cross section (NRCS) of the glasses involved. The results showed that Ag2O and WO3 contents had an insignificant effect on the mechanical properties of the investigated glass system. In contrast, the Ag2O and WO3 contents had a significant effect on the shielding capacity of the glasses against gamma radiation and charged particles. The Zeff values were in the range of 24–60 for gamma, 24–31 for electron, 18–24 for proton, 15–23 for alpha, and 13–19 for carbon ion. The shielding properties of these glasses were compared with those of commonly used shields. The investigated glass system can be used in shielding applications, wherein Ag2O and WO3 contents can be balanced with the TeO2 content according to the type and energy of the radiation.

A mini subcritical nuclear reactor was designed using Monte Carlo methods. The reactor has light water as moderator, natural uranium as fuel, and a 239PuBe neutron source. In the design uranium fuel was modeled in an arrangement of concentric rings: 8.5, 14.5, 20.5 26.5, 32.5 cm-inner radius, 3 cm-thick, and 36 cm-high. Different models were made from a single ring of natural uranium to five rings. For each case, the neutron spectra, the neutron fluence distribution, the effective multiplication factor, the amplification factor, and the reactor power were estimated. The ambient dose equivalent rate outside the mini reactor was also estimated. The maximum value for the keff (0.78) was obtained when five rings of fuel were used; this value is close to 0.86 which belongs to a Nuclear Chicago subcritical reactor which requires almost twice the amount of uranium than the mini subcritical reactor.

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

This paper examines the gamma-ray shielding features of some selected tellurite-based glasses in the form of xMoO3(100−x)TeO2 (20 ≤ x ≤ 50 mol%). Mass attenuation coefficients (μ m ) of these glasses have been calculated using Geant4 toolkit and XCOM program for photon energy range of 1 keV—1000 MeV. The correlation factor (R2) between these two methods was found to be almost one. Shielding effectiveness for these glasses has been estimated by computing half value layer (HVL), effective atomic number (Z eff ), and mean free paths (MFP). It was noticed that the heavy metal oxide of TeO2 plays an important role in improving the shielding effectiveness of the glasses. The Mo20Te80 glass has shown the promising properties to serve for gamma ray protection applications as compared with many conventional concretes and other newly developed glasses.

The mass attenuation coefficients of seven different types of normal and heavy concretes like ordinary, hematite-serpentine, ilmenite-limonite, basalt-magnetite, ilmenite, steel-scrap and steel-magnetite concretes has been simulated using FLUKA Monte Carlo code at high energies 1.5, 2, 3, 4, 5 and 6 MeV. The mass attenuation coefficients and linear attenuation coefficient of the concretes were found dependent upon the chemical composition, density and gamma ray energy. FLUKA Monte Carlo code results were found in good agreement with experimental and theoretical XCOM data. Our investigations for high energy gamma-ray interaction validate the FLUKA Monte Carlo code for use where experimental gamma-ray interaction results are not available.

The shielding characteristics of quarry tuff (cantera) against photons and 241 AmBe neutrons, were determined. The effective atomic number of cantera, the Exposure and the Energy absorption buildup factors for photons in cantera were also calculated. The XCOM code was used to calculate the photon interaction coefficients. Also, Monte Carlo method was used to model a photon transmission experiment in cantera. Collided and uncollided photon fluence, Kerma in air, and Ambient dose equivalent were estimated. With the uncollided photon fluence the linear attenuation coefficients were determined and compared with those calculated with the XCOM code. The linear attenuation coefficient for 0.662 MeV photons was compared with the coefficient measured with a NaI(Tl)-based γ-ray spectrometer and a 137 Cs source. The Monte Carlo model was also used to estimate the neutron spectra of 241 AmBe neutrons in function of cantera thickness, the Effective and the Ambient dose equivalent for the collided and uncollided neutrons. Cantera has good shielding properties for low energy photons and poor shielding features against 241 AmBe neutrons.

When carrying out experimental work related to neutron interactions with the matter, it is important to get an accurate knowledge about the angular distributions of neutron beam. Descriptions of the neutron source assembly as well as the irradiator design are discussed. A sealed tube neutron generator is used to produce 14 MeV neutrons through deuterium (2H) - tritium (3H) reactions. Monte Carlo simulations have been carried out for optimizing 14 MeV neutron flux distribution around a tritium target. Simulation was also tested by experimental measurements using a foil activation method. The discrepancy between two methods is due to sample preparations, irradiation facility and geometry.

A series of ternary phosphate glasses in the form of (PbO)x(ZnO)60-x(P2O5)40 where x = 0–60 mol%, have been successfully prepared by conventional melt-quenching technique. The physical and elastic properties of the glasses have been investigated using pulse echo technique. The longitudinal and shear velocity of the glasses were measured using the MBS8000 ultrasonic data acquisition system at 10 MHz frequency in room temperature. The density, ultrasonic velocity and elastic moduli are found to be composition dependent and the correlation between the elastic moduli with the atomic packing density is discussed in detailed. The shielding parameters, mass attenuation coefficients, half value layers and exposure buildup factor (EBF) values have been computed using WinXCom program with the use of GP fitting method, and variation of shielding parameters are discussed for the effect of PbO addition into the glasses and photon energy. An increase in the density of the glasses results in a change in crosslink density. The sound velocity and elastic properties increased with PbO content and increase in Poisson's ratio trend suggests that the rigidity of the glasses has decreased. Besides, the replacement of ZnO by PbO causes an increase in mass attenuation coefficient, while the half value layer and the exposure buildup factor were decreased and these glasses has been potentially used as shielding material.

Mass attenuation coefficients ([Formula: see text]) for some nonlinear optical materials such as potassium dihydrogen phosphate, ammonium dihydrogen phosphate, zinc tris-thiourea sulphate, and zinc thiourea chloride were measured using a [Formula: see text] NaI(Tl) scintillation detector at gamma energies of 122 keV, 356 keV, 511 keV, 662 keV, 840 keV, 1170 keV, 1270 keV, and 1330 keV. In addition, GEANT4 simulations were carried out to mimic the experiment at these energies. As a result, good agreement between the experimental and GEANT4 results was observed. The measured [Formula: see text] values were used to compute effective atomic numbers ([Formula: see text]) for the selected materials. It was found that the [Formula: see text] values were in the range typical for dosimetric materials.

In this study, 12 different concentrations of shielding materials were developed and produced. They were covered with high temperature resistant (1500 degrees 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 Am-241-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., Project of Agri Ibrahim Cecen University Scientific Research Projects [MYO 18.001-19.001], This study was supported by the Project of Agri Ibrahim Cecen University Scientific Research Projects (MYO 18.001-19.001).

A series of ultraviolet B (UVB, 280–315 nm) radiation emitting Gd doped Sr3P4O13 was synthesized using the co-precipitation method. Formation of products was confirmed by X-ray diffraction analysis, while scanning electron microscope was employed to investigate the morphology of the prepared product. The emission spectra of the prepared samples were recorded at 272 nm excitation wavelength with different Gd3+ concentrations (0.005–0.15 mol). Electron paramagnetic resonance spectra of the Gd3+ doped samples showed the substitution of Sr2+ sites by Gd3+ ions with lower symmetry at X-band frequency. Shielding parameters for all the un-doped and Gd doped Sr3P4O13 samples were determined in the range 1 keV–100 GeV photon energy. These photon interaction parameters vary with the compositions and the photon energy. These parameters were found to be maximum for the gadolinium concentration 0.15 mol. The current investigation and obtained results would be helpful for shielding evaluation of radioactive waste containment in the nuclear fuel cycle.

In this study, the mass attenuation coefficient of boron-containing ores in the Liaoning province of China was calculated using WinXCOM software to investigate the shielding effectiveness of these ores against gamma rays. The mass attenuation coefficients were also calculated using MCNP-4B code and compared with WinXCOM results; consequently, a good consistency between the results of WinXCOM and MCNP-4B was observed. Furthermore, the G-P fitting method was used to evaluate the values of exposure buildup factor (EBF) in the energy range of 0.015–15 MeV up to 40 mean free paths. Among the selected ores, boron-bearing iron concentrate ore (M3) was determined to be the best gamma ray shielding ore owing to its higher values of mass attenuation coefficient and equivalent atomic number and lower value of EBF. Moreover, American Evaluated Nuclear Data File (ENDF/B-VII) was used to analyze the shielding effectiveness against thermal neutrons. It was determined that Szaibelyite (M2) is the best thermal neutron shielding material. This study would be useful for demonstrating the potential of boron-containing ores for applications in the field of nuclear engineering and technology.

Mass attenuation coefficient, μ/ρ for some polymers with potential applications in dosimetry, and medical physics has been calculated using Monte Carlo simulation code, MCNP-4C at different gamma ray energies, 59.5, 279.1, 511, 661.6, 662, 1115.5, 1173, 1173.2, 1274.5 and 1332.5 keV. Appreciable variations were noted for mass attenuation coefficients for the polymers by changing the photon energy. The simulated μ/ρ values were compared with possible available experimental data and theoretical XCOM results, and good agreement was observed. Present study indicates that simulation geometry method is suitable to be used as an alternative method for the experiments. The present geometry can be used as standard geometry for MCNP simulation for low-Z materials. The calculated μ/ρ values using MCNP-4C code signify that the simulation geometry method can be followed to determine the gamma ray attenuation coefficients for the polymers for which there are no experimental values available.

The Kerma factors and Kerma relative to the air and water of 24 compounds that are used as tissue substitutes were calculated for neutron energies ranging from 2.53×10(-8) to 29 MeV. The Kerma ratios of the tissue substitutes relative to air and water were calculated. The water equivalence of the selected tissue substitutes was observed above neutron energies of 100 eV. The Kerma ratio relative to the air for poly-vinylidene fluoride and Teflon were nearest to unity at very low energy (up to 1 eV) and above 63 eV, respectively. It was found that the natural rubber was a water-equivalent tissue substitute compound. The results of the Kerma factors in our investigation show good agreement with those published in ICRU-44. We found that at higher neutron energies, the Kerma factors and Kerma ratios of the selected tissue substitute compounds were approximately the same, but though the differences were large for energies below 100 eV.

Gamma ray exposure buildup factor (EBF) and neutron shielding efficiency of borate glass systems in composition (100 − x − y) Na2B4O7−xPbO−yNiO (where x and y = 0, 2, 4, 6, 8 and 10 weight percentage) were investigated in the present study. Gamma ray EBF values were computed using Geometric Progression (G-P) method for photon energy 0.015 to 15 MeV up to penetration depths of 40 mfp (mean free path). The macroscopic effective removal cross-sections (Σ R ) for fast neutron (2 to 12 MeV) were calculated. At low- and high-energy region, the EBF values were found minimum whereas maximum in the intermediate-energy regions with a sharp peak below 100 keV for oxide containing glass samples. The investigation was carried out for potential applications of borate glasses for radiation shielding.

In the present work, we computed the γ-ray interaction characteristics of some boron containing materials such as B 3 Al 2 O 3 , B 4 C, B 10 H 14 and ferro-boron (Fe–B) by means of effective atomic numbers and exposure buildup factors (EBF). The EBF of the materials were calculated using Geometric Progression (G-P) fitting for photon energy range 0.015–15 MeV up to penetration depth of 40 mean free path. The Fe–B is found to be the superior γ-ray shielding material. The results of the present work could be useful in radiation shielding applications in various neutron sources, reactors and accelerators.

Gamma ray shielding effectiveness of superconductors with a high mass density has been investigated. We calculated the mass attenuation coefficients, the mean free path (mfp) and the exposure buildup factor (EBF). The gamma ray EBF was computed using the Geometric Progression (G-P) fitting method at energies 0.015-15 MeV, and for penetration depths up to 40 mfp. The fast-neutron shielding effectiveness has been characterized by the effective neutron removal cross-section of the superconductors. It is shown that CaPtSi3, CaIrSi3, and Bi2Sr2Ca1Cu2O8.2 are superior shielding materials for gamma rays and Tl0.6Rb0.4Fe1.67Se2 for fast neutrons. The present work should be useful in various applications of superconductors in fusion engineering and design. (C) 2014 Elsevier Ltd. All rights reserved.

In the proposed work, an investigation on hazard assessment by lead element and natural radioactivity in cosmetic samples collected from various countries is presented. These samples were face powder, eyebrow paint and henna. The lead element in cosmetic samples was determined using particle-induced X-ray emission. Maximum natural radioactivity concentrations of 226Ra and 40K were found in khol and make-up cosmetic samples, respectively. The qualitative analysis of cosmetic samples showed that lead is the most toxic element found in eyebrow paint samples.

Alloys, concretes, glasses, neutron shielding material, polymers, nuclear track detectors, are being used for shielding and dosimetry applications. The neutron kerma coefficients of eighteen compounds have been calculated for neutrons energy less than 30 MeV; these compounds are used as shielding and dosimetric materials. The kerma coefficients are found lowest for all the compounds in intermediate energy (10 eV E Z elements. A resonance in neutron kerma coefficient is observed for the large amount of oxygen containing compounds. It is found that at high energies kerma coefficients are approximately same, but very large differences for energy less than 100 eV.

Geant4 Monte Carlo code simulations were used to solve experimental and theoretical complications for calculation of mass energy-absorption coefficients of elements, air, and compounds. The mass energy-absorption coefficients for nuclear track detectors were computed first time using Geant4 Monte Carlo code for energy 1 keV-20 MeV. Very good agreements for simulated results of mass energy-absorption coefficients for carbon, nitrogen, silicon, sodium iodide and nuclear track detectors were observed on comparison with the values reported in the literatures. Kerma relative to air for energy 1 keV-20 MeV and energy absorption buildup factors for energy 50 keV-10 MeV up to 10 mfp penetration depths of the selected nuclear track detectors were also calculated to evaluate the absorption of the gamma photons. Geant4 simulation can be utilized for estimation of mass energy-absorption coefficients in elements and composite materials. (C) 2014 Elsevier Ltd. All rights reserved.

The mass attenuation coefficients, mu/rho and effective atomic numbers, Z(eff) of some carbon steel and stainless steel alloys have been calculated by using Geant4, MCNP simulation codes for different gamma ray energies, 279.1 keV, 661.6 keV, 662 keV, 1115.5 keV, 1173 keV and 1332 key. The simulation results of Zeff using Geant4 and MCNP codes have been compared with possible available experimental results and theoretical WinXcom, and good agreement has been observed. The simulated pip and Zeff values using Geant4 and MCNP code signifies that both the simulation process can be followed to determine the gamma ray interaction properties of the alloys for energies wherever analogous experimental results may not be available. This kind of studies can be used for various applications such as for radiation dosimetry, medical and radiation shielding. (C) 2014 Elsevier Ltd. All rights reserved.

The mass attenuation coefficients of some low-Z dosimetric materials with potential applications in dosimetry, medical and radiation protection have been investigated using the Monte Carlo simulation code Monte Carlo N-Particle (MCNP). Appreciable variations are noted for the mass attenuation coefficient by changing the photon energy. The MCNP-simulated parameters are compared with the experimental data wherever possible and theoretical values through the WinXcom program. The simulated results obtained by MCNP generally agree well with the experiment and WinXcom predictions for various low-Z dosimetric and tissue substitute materials. In addition, the mass attenuation coefficients around the k-edges for low-Z dosimetric materials estimated from the MCNP code agree very well with WinXcom prediction. Finally, the results indicate that this simulation process can be followed to determine the interaction parameters of gamma rays in such low-Z materials for which there are no satisfactory experimental values a...

Optically stimulated luminescence (OSL) materials are sensitive dosimetric materials used for precise and accurate dose measurement for low-energy ionizing radiation. Low dose measurement capability with improved sensitivity makes these dosimeters very useful for diagnostic imaging, personnel monitoring and environmental radiation dosimetry. Gamma ray energy absorption buildup factors and exposure build factors were computed for OSL materials using the five-parameter Geometric Progression (G-P) fitting method in the energy range 0.015–15 MeV for penetration depths up to 40 mean free path. The computed energy absorption buildup factor and exposure buildup factor values were studied as a function of penetration depth and incident photon energy. Effective atomic numbers and Kerma relative to air of the selected OSL materials and tissue equivalence were computed and compared with that of water, PMMA and ICRU standard tissues. The buildup factors and kerma relative to air were found dependent upon effective atomic numbers. Buildup factors determined in the present work should be useful in radiation dosimetry, medical diagnostics and therapy, space dosimetry, accident dosimetry and personnel monitoring.

Gamma-ray shielding competence of silicate and borate heavy metal oxide glasses has been investigated using linear attenuation coefficients, effective atomic numbers and exposure buildup factors (EBF). The gamma-ray EBF were computed using the Geometric Progression (G-P) fitting method for photon energies from 0.015 to 15 MeV, and for penetration depths up to 40 mean free paths (mfps). The macroscopic effective removal cross-section for fast neutron has been calculated for energy range from 2 to 12 MeV. It is found that bismuth silicate glass has superior shielding properties and is suitable for replacement of lead glasses. The present investigation is very useful for gamma-ray and neutron shielding and design for lead-free shielding glass in radiological engineering.

Monte Carlo N-particle (MCNP) code has been used to simulate the transport of gamma photon rays of different energies (22, 31, 59.5 and 81 keV) to estimate the iron content in solutions. In this study, MCNP simulation results are compared with experiment and XCOM theoretical data. The simulation shows that the obtained results are in good agreement with experimental data, and better than the theoretical XCOM values. The study indicates that MCNP simulation is an excellent tool to estimate the iron concentration in the blood samples. The MCNP code can also be utilized to estimate other trace elements in the blood samples.

The gel dosimeter comprises of phantom, dosimetric material and three-D spatial dose distribution has advantages over one- and two-D dosimeters. Energy absorption buildup factor (EABF) and exposure buildup factor (EBF) values of sixteen gel dosimeters have been computed for photon energy 0.015 to 15 MeV up to 40 mfp (mean free path) penetration depths. Kerma of the gel dosimeters were computed for photon energy 1 keV to 20 MeV. The water and PMMA phantom equivalence of the gel dosimeters was evaluated using EABF, and large difference was noted below 1 MeV photon energy. This study should be useful for estimation of effective dose to the human organs and simulation of the dose for radiation therapy and various medical applications.

The main goal of this present study is focused on testing the applicability of Geant4 electromagnetic models for studying mass attenuations coefficients for different types of composite materials at 59.5, 80, 356, 661.6, 1173.2 and 1332.5keV photon energies. The simulated results of mass attenuation coefficients were compared with the experimental and theoretical XCOM data for the same samples and a good agreement has been observed. The results indicate that this process can be followed to determine the data on the attenuation of gamma rays with the several energies in different materials. The modeling for photon interaction parameters was standard for any type of composite samples. The Geant4 code can be utilized for gamma ray attenuation coefficients for the sample at different energies, which may sometimes be impractical by experiment investigation.

In the present work, different methods were used to determine the effective atomic numbers of some low-Z materials, namely, polyethylene (PE), polystyrene (PS), polypropylene (PP), Perspex (PX), polycarbonate (PC), nylon 6-6 (PA-6), plaster of Paris (POP), and TH/L2. These methods are the direct method, the interpolation method, Auto-Zeff software, and single value XMuDat computer program. Some of the results obtained were compared with experimental data wherever possible. It can be concluded from this work that the effective atomic numbers calculated with the direct, the interpolation and Auto-Zeff methods demonstrate a good agreement in Compton scattering and pair production energy regions. A large difference in the effective atomic numbers calculated by the direct and the interpolation methods of low-Z materials was also observed in photoelectric and pair production regions. It was determined that PE, PS, PX, and PA-6 were equivalent to adipose and muscle; POP was equivalent to cortical bone; TH/L2 was equivalent to thyroid tissue; PP was equivalent to yellow bone marrow and adipose tissues; PC was equivalent to spongiosa.

The mass attenuation coefficients, mu/rho of some thermoluminescent dosimetric materials with potential applications as personnel dosimeters, have been investigated using Geant4 simulation toolkit. Appreciable variations are noted for the mu/rho values by changing the photon energy and chemical compositions. The Geant4 toolkit simulation results of mu/rho are compared with experimental data wherever possible. Comparisons are also made with predictions from the XCOM program in the energy region from I keV to 100 GeV. Good agreement among Geant4 toolkit simulation, XCOM program and experimental data indicate that the Geant4 toolkit may be employed to make calculations on the mu/rho values of various dosimetric materials. The results indicate that this process can be followed to determine the interaction of photon with the several energies in different materials. (C) 2014 Elsevier Ltd. All rights reserved.

The gamma-ray shielding effectiveness of some oxide dispersion-strengthen (ODS) alloys by means of mass attenuation coefficients, mean free path, exposure buildup factors and energy absorption buildup factors have been investigated in the present study. The buildup factors were calculated using geometrical progression method for photon energy 0.015–15 MeV up to 40 mfp penetration depth. The mass attenuation coefficients were calculated by using XCOM program and Geant4 simulation methods and found a very good agreement. Our investigation signifies that the low iron content ODS alloys are superior shielding materials with the lower buildup factors. This study should be useful for selection of shielding materials for their applications in fusion reactors design and future nuclear reactor technologies.

Gamma-ray exposure buildup factor (EBF) values and neutron shielding effectiveness of bismuth borosilicate (BBS) glass systems in composition (50− x )SiO 2 :15B 2 O 3 :2Al 2 O 3 :10CaO:23Na 2 O: x Bi 2 O 3 (where x =0, 5, 10, 15 and 20 mol%) were calculated. The EBF values were computed for photon energy 0.015–15 MeV up to penetration depths of 40 mfp (mean free path) by the geometrical progression (G-P) method. The EBF values were found dependent upon incident photon energy, penetration and bismuth molar concentration. In low- and high-energy photon regions, the EBF values were minimum whereas maximum in the intermediate-energy region. The fast neutron removal cross sections for energy 2–12 MeV were calculated by the partial density method. The BBS glass with 20 mol% Bi 2 O 3 is found to be superior gamma-ray and neutron transparent shielding. The EBF values of the BBS glasses were compared with steel–magnetite concrete and lead. The investigation was carried out to explore the advantages of the BBS glasses in different radiation shielding applications.

We have computed γ-ray exposure buildup factors (EBF) of some building materials; glass, marble, flyash, cement, limestone, brick, plaster of paris (POP) and gypsum for energy 0.015–15 MeV up to 40 mfp (mfp, mean free path) penetration depth. Also, the macroscopic effective removal cross-sections (ΣR) for fast neutron were calculated. We discussed the dependency of EBF values on photon energy, penetration depth and chemical elements. The half-value layer and kinetic energy per unit mass relative to air of building materials were calculated for assessment of shielding effectiveness. Shielding thicknesses for glass, marble, flyash, cement, limestone and gypsum plaster (or Plaster of Paris, POP) were found comparable with ordinary concrete. Among the studied materials limestone and POP showed superior shielding properties for γ-ray and neutron, respectively. Radiation safety inside houses, schools and primary health centers for sheltering and annual dose can be assessed by the determination of shielding para...

The gamma-ray exposure buildup factors of oxide dispersive-strengthen (ODS) steel alloys have been computed for photon energy 0.015–15 MeV up to 40 mean free path (mfp) penetration depth by geometrical progression (GP) method. An appreciable variation in exposure buildup factors for the investigated ODS steel alloys was observed. The exposure buildup factors of ODS alloys were observed strongly depending upon the photon energy, penetration depths and the chemical compositions. The exposure buildup factors were found to be small in low- and high-photon energy whereas these were very large in intermediate-energy region. Our investigation signifies that the Eurofer 97 ODS steel alloy is the superior gamma-ray and neutron shielding materials. This study should be useful for shielding design and application in future reactor technologies and fusion reactor blankets.

The effective atomic weight, A eff , mass attenuation coefficients, μ/ρ, half-value layer, HVL, effective atomic numbers, Z eff and effective electron densities, N eff of hydride and borohydirde metals, with potential shielding applications in fusion reactors have been investigated in the present work. The gamma ray interaction parameters, μ/ρ, HVL, Z eff and N eff were calculated for photon energy range 1 keV–100 GeV. The A eff was calculated using fast neutron removal cross-sections (Σ R ) for neutron energy 2–12 MeV. The effective atomic number using fast neutron Σ R was evaluated for compound/composite materials first time. The compounds or mixture containing low as well high atomic weight elements are required for fast neutron shielding. The Mg (BH4)2, TiH2 and ZrH2 hydride and borohydride metals are found to be the superior shielding materials. This study should be useful for design of effective shielding using hydrides and borohydrides metals in fusion reactors.

The mass attenuation coefficients, l/q, the linear attenuation coefficients, l and the mean free path, MFP of normal and heavy concretes; ordinary, hematite- serpentine, ilmenite-limonite, basalt-magnetite, ilmenite, steel-scrap and steel-magnetite have been investigated using Geant4 Monte Carlo method at photon energies 1.5, 2, 3, 4, 5 and 6 MeV. The mass attenuation coefficients, the linear attenuation coefficient and the mean free path for the concretes were found dependent upon chemical composi- tions, density and the photon energy. The linear attenuation coefficient values for the selected concretes increase with the density and decrease with the energy. The mean free path thickness for the all the concretes decrease with increase in iron content and increase with increase the photon energy. The Geant4 Monte Carlo simulation method results have been compared with experimental and theoretical XCOM data, and showed a very good agree- ment. Our investigation validates the utilization of the Geant4 Monte Carlo simulation method for high energy photon interactions.

Gamma ray and neutron shielding properties of some alloy materials, CS-516, SS-403, SS-410, SS-316, SS-316L, SS-304L, Incoloy-600, Monel-400 and Cupero-Nickel were studied in the present work. Gamma shielding effectiveness of the alloys was studied by calculation of mass attenuation coefficient, half-value layer and exposure buildup factor (EBF). The EBFs of the alloys were calculated by GP fitting formula for photon energy 0.015–15 MeV up to 40 mfp penetration depth. Fast neutron removal cross section of the alloys was calculated by partial density method. Cupero-Nickel was found to be best shielding for gamma rays. For neutron, SS-316 was found to be the best shielding materials in energy 2–12 MeV. This study should be useful for potential applications of these materials in nuclear reactor core design and other industries for choice of effective gamma ray and neutron shielding materials.

Shielding of a reactor is required for protection of people and environment during normal operation and accidental situations. In the present paper we investigated the shielding parameters viz. mass attenuation coefficients, linear attenuation coefficients, tenth-value layer, effective atomic numbers, kerma relative to air and exposure buildup factors for gamma-ray for ordinary, heavy, and super heavy concretes. Macroscopic effective removal cross-sections for fast neutron had also been calculated. Ordinary concrete is economically suitable for mixture high energy gamma-ray and neutron as it has large weight fraction of low-Z as compared with super heavy concretes to slow down the neutron. Super heavy concretes are superior shielding for both reactor operation and accident situations. The study is useful for optimizing for shielding design and radiation protection in the reactors.

The mass attenuation coefficients, μ / ρ , total interaction cross-section, σ t , and mean free path ( MFP ) of some Heavy Metal Oxides (HMO) glasses, with potential applications as gamma ray shielding materials, have been investigated using the MCNP-4C code. Appreciable variations are noted for all parameters by changing the photon energy and the chemical composition of HMO glasses. The numerical simulations parameters are compared with experimental data wherever possible. Comparisons are also made with predictions from the XCOM program in the energy region from 1 keV to 100 MeV. Good agreement noticed indicates that the chosen Monte Carlo method may be employed to make additional calculations on the photon attenuation characteristics of different glass systems, a capability particularly useful in cases where no analogous experimental data exist.

The mass attenuation coefficients, linear attenuation coefficients and tenth value layer of seven types of concretes having density (2.3–5.11 g cm−3) have been simulated by Monte Carlo simulation techniques at photon energies 1.5, 2, 3, 4, 5 and 6 MeV. Numerical simulation results have been compared with experimental and XCOM photon cross section database and showed a good agreement. Linear attenuation coefficient and tenth value layer of the concretes were found chemical composition, density and photon energy dependent. The tenth value layer of the concretes decrease with increase iron content and increase with increase in photon energy. The linear attenuation coefficients values of the concretes increase with density and decrease with photon energy. Monte Carlo simulation, calculated and experimental results show that the ordinary concrete having lowest iron content (0.63%) have minimum linear attenuation coefficient.

A comprehensive study of gamma-ray exposure build-up factors (EBFs) of fly-ash brick materials has been carried out for photon energies of 0.015–15 MeV up to a penetration depth of 40 mfp (mean free path) by a geometrical progression (GP) fitting method. The EBF values of the fly-ash brick materials were found to be dependent upon the photon energy, penetration depth and chemical composition, and were found to be higher than the values for mud bricks and common bricks. Above a photon energy of 3 MeV for large penetration depths (>10 mfp), the EBF becomes directly proportional to Zeq. EBFs of fly-ashes were found to be less than or equal to those of concrete for low penetration depths (

The gamma-ray exposure build-up factors of raw materials of bricks (soils and fly-ashes) in the state of Punjab were investigated for the photon energy range 0.015 to 15 MeV up to 40 mfp penetration depth by the geometrical progression (GP) method. Appreciable variations in the exposure build-up factor (EBF) are noted for the raw materials. The EBFs of the raw materials of bricks change depending on the photon energy, penetration depth and chemical composition. The build-up factors are low at low and high photon energies, whereas they are very high in the medium-energy region. The peak energy of the EBF for soils is 0.3 MeV and 0.2 MeV for fly ashes. The EBFs of the raw materials of bricks are also compared with those of bricks of red mud and common brick materials. Common bricks were found to have the lowest gamma-ray EBF. This study should be useful for emergency preparedness planning and emergency dose estimation for future planned nuclear power plants in the state of Punjab.

The mass attenuation coefficient (µ/ρ), effective atomic number (ZPleff ) and effective electron density (Neff ) of carbon steel and stainless steels were computed by using the WinXcom program. The µ/ρ values are higher in the photo-electric absorption and pair production region and approximately constant in the Compton scattering region. The variation in µ/ρ values is explained by partial interaction processes. The effective atomic numbers of the steels were also calculated by the Auto-Zeff program. The ZPleff values of the steels were compared with experimental results for available energies and were found to be in good agreement. SS304L shows equal values of effective atomic number by both the methods in the region of interest. The electron densities of the selected steels are observed to be equal and approximately constant in the Compton scattering region.

Effective atomic numbers (ZPIeff) and electron densities of eighteen alcohols such as wood alcohol, CH3OH; grain alcohol, C2H5OH; rubbing alcohol, C3H7OH; butanol, C4H9OH; amyl alcohol, C5H11OH; cetyl alcohol, C16H33OH; ethylene glycol, C2H4(OH)2; glycerin, C3H5(OH)3; PVA, C2H4O; erythritol, C4H6(OH)4; xylitol, C5H7(OH)5; sorbitol, C6H8(OH)6; volemitol, C7H9(OH)7; allyl alcohol, C3H5OH; geraniol, C10H17OH; propargyl alcohol, C3H3OH; inositol, C6H6(OH)6, and menthol, C10H19OH have been calculated in the photon energy region of 1 keV-100 GeV. The estimated values have been compared with experimental values wherever possible. The comparison of ZPIeff of the alcohols with water phantom and PMMA phantom indicate that the ethylene glycol, glycerin, and PVA are substitute for PMMA phantom and PVA is substitute of water phantom. ZPIeff of alcohols have also been compared with human organs and tissues. Ethylene glycol, glycerin and PVA, allyl alcohol, and wood alcohols are found tissue substitutes for most of human organs. Kerma which is the product of the energy fluence and mass energy-absorption coefficient, have been calculated in the energy region from 1 keV to 20 MeV for the alcohols. The results show the kerma is more or less independent of energy above 100 keV.

Gamma-ray energy absorption buildup factors of gaseous mixtures; neon (95%) + argon (5%), argon (95%) + acetylene (5%), argon (95%) + methane (5%), argon (95%) + carbon dioxide (5%), methane (70%) + pentane (30%) and argon (90%) + methane (10%) were studied by Geometrical Progression (G-P) fitting for the photon energy range 0.015-15 MeV. It was found that the equivalent atomic number, Z eq of the gaseous mixtures sharply reduces after 1 MeV photon energy. The Z eq for the mixture of methane (70%) + pentane (30%) is the minimum, whereas the maximum is for argon (95%) + carbon dioxide (5%) for the photon energies under investigation. The Energy Absorption Buildup Factor (EABF) for methane (70%) + pentane (30%) was found to be the highest among all the selected gaseous mixtures. The chemical composition of the gaseous mixtures has an impact on the EABF values for photon energy and penetration depth. The investigation of the EABF is useful for selection of gaseous mixtures in design consideration of gaseous radiation detectors for gamma radiation.

Investigation of Nanoparticle Effect on Radiation Shielding Property Using Monte Carlo Method Radiation protection and choice of appropriate materials in the areas of medical and industrial applications has become one of the major research topics. In this study, the increment in mass attenuation coefficients by doping of nano-sized barite (BaSO4) into the lead (Pb) have been investigated. The validation of generated MCNPX simulation geometry is provided by comparison with the previous experimental and theoretical studies. This generated geometry was used for the definition of nano-BaSO 4 into lead material and calculations. It was found that nano-BaSO 4 doped lead increased the amount of radiation absorbed in the material inside and as a result of which affected the radiation attenuation properties of material. This study also showed that MCNPX is an effective code on nano size studies and standarised geometry can be useful for further investigations. Radyasyon Zirhlama Ozelliginde Nanoparcacik Etkisinin Monte Carlo Metodu Kullanarak Incelenmesi Radyasyondan korunma ve uygun malzeme secimi medikal ve endustriyel radyasyonun kullanildigi alanlarda baslica arastirma konularindan bir tanesi haline gelmistir. Bu calismada, kursun (Pb) icerisine nano boyutlarda barit (BaSO 4 ) katkisi yapilarak kutle zayiflatma katsayisinda meydana getirdigi artis incelenmistir. Olusturulan MCNPX simulasyon geometrisinin validasyonu onceki deneysel ve teorik olcumler ile saglanmistir. Olusturulan bu geometri nano BaSO 4 parcaciklarinin kursun malzeme icerisine tanimlanmasinda ve hesaplamalarda kullanilmistir. Sonuc olarak nano boyutlarda kursun icerisine katkilanan BaSO 4 nanoparcaciklarikursunun kutle zayiflatma katsayisinda bir artisa sebep olmus, malzeme icerinde sogurulan radyasyon miktari artmis ve malzemenin radyasyonun zayiflatma ozelliklerini etkilemistir. Bu calisma ayrica MCNPX kodunun nano boyutlardaki calismalarda etkili bir program oldugunu ve standartize edilen bu geometrinin gelecek calismalar icin kullanilabilirligini gostermistir.