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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.

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.

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 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.

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.

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.

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.

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.

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.

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.

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.

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 radiation shielding efficiency of material depends upon photon attenuation, exposure buildup factors and neutron removal capacity. A newly developed Pb-free gadolinium-based glasses in compositions (80-x) B2O3-10SiO2-10CaO-xGd2O3 (where x = 15, 20, 25, 30 and 35 mol%) had completely been investigated for their shielding efficiency with Geant4 simulation for mass attenuation coefficients and neutron total macroscopic cross section and by calculating exposure buildup factors. The exposure buildup factors for photon energy from 0.015 to 15 MeV had been calculated up to 40 mean free paths using five factors geometric progression method. The mass attenuation coefficients of the Pb-free glasses were simulated for energies from 223 to 2614 keV and compared with the possible available experimental results. The neutron shielding efficiency of these glasses was discussed by calculating neutron total macroscopic cross section for energies from 1 eV to 14.1 MeV. Present investigations are found to be very useful for applications in nuclear engineering.

This study aimed to determine the gamma shielding parameters of an Erbium Zinc Tellurite Glass System. The mass attenuation coefficients of Erbium Zinc Tellurite Glass System xEr2O3:20ZnO : (80-x) TeO2 were computed using Monte Carlo simulation at photon energies 20, 30, and 40 keV and 60, 356, 662, 1,173, 1,274, and 1,332 keV. The half-value layer thickness was calculated using the mass attenuation coefficients. The obtained results were compared to standard XCOM data and found comparable. It was found that the erbium zinc tellurite glass with a higher content of erbium is superior shielding material to other glasses. Thus, the validation of a generated Monte Carlo N-Particle Transport Code System-Extended simulation setup has been provided. It can be concluded that results of the present investigation would be very useful for applications of medical diagnostic and nuclear technology.

The mass attenuation coefficients, μ/ρ, half-value layer, HVL, tenth-value layer, TVL, effective atomic numbers, ZPIeff, and effective electron densities, Ne,eff, of borate glass sample systems of (100-x-y) Na2B4O7 : xPbO : yNiO (where x and y=0, 2, 4, 6, 8, and 10 weight percentage) containing PbO and NiO, with potential gamma ray and neutron shielding applications, have been investigated. The gamma ray interaction parameters, μ/ρ, HVL, TVL, ZPIeff, and Ne,eff, were computed for photon energy range 1 keV–100 GeV. The macroscopic fast neutron removal cross-sections (ΣR) have also been calculated. Appreciable variations were noted for all the interaction parameters by varying the photon energy and the chemical composition of the glass samples. The better shielding properties of borate glass samples containing PbO were found. These results indicated that borate glass samples are a good radiation shielding material.

Gamma ray build-up factors of concrete shielding have been investigated by Geometrical Progression fitting formula for photon energy range 0.015–15 MeV up to 40 mfp penetration depth. The energy absorption build-up factors and Exposure Build-Up Factor (EBF) of concrete shielding have been studied as a function of photon energy, penetration depth, concrete density and chemical element compositions. The EBF of the concrete shielding having high equivalent atomic number increases rapidly beyond 3 MeV photon energy for high penetration depths. The build-up factors show that the heavy concretes (≥ 3.5 g cm–3) are good biological shielding. Steel-magnetite (5.11 g cm–3) is the best shielding concrete for reactor core for power operation as well as accidents. Present build-up factors are very useful for selection of shielding materials for reactor core in controlling the gamma exposure to the people involved and the public of photon energy abundance in the range of 0.10–10 MeV.

A simple and economical passive graded shielding which is combination of low and high Z materials for HPGe detector for background radiation minimisation is developed for better identification of the peaks in acquired spectrum by reducing Compton continuum. The present paper discusses the design and advantages of the passive graded shielding (lead + copper + aluminium) which encloses the HPGe detector. The passive graded shielding improved sensitivity of the detector and reduced Minimum Detectable Activity (MDA) for planchet source from 19.6% to 11.7% for gamma photon energies in the range of 81–1408 keV and brought down the MDA, 35.25 GBq d –1 and 35.26 GBq d –1 for 133 Xe and 135 Xe respectively which was 72.31 GBq d –1 and 78.27 GBq d –1 for noble gases in gaseous effluents before shielding of the detector. The passive graded shielding as integral part is considerably very useful for radiological protection at nuclear power/research reactors by minimising background radiation.