Yazar "Singh, V. P." seçeneğine göre listele

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    Öğe
    Comparison of mass attenuation coefficients of concretes using FLUKA, XCOM and experiment results
    (Edp Sciences S A, 2018) Singh, V. P.; Korkut, T.; Badiger, N. M.
    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 6MeV. 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 gammaray interaction validate the FLUKA Monte Carlo code for use where experimental gamma-ray interaction results are not available.
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    Öğe
    Experimental and Monte Carlo simulation study on potential new composite materials to moderate neutron-gamma radiation
    (Pergamon-Elsevier Science Ltd, 2020) Aygun, Bunyamin; Sakar, Erdem; Singh, V. P.; Sayyed, M. I.; Korkut, Turgay; Karabulut, Abdulhalik
    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.
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    Öğe
    Gamma exposure buildup factors and neutron total cross section of ceramic hosts for high level radioactive wastes
    (Pergamon-Elsevier Science Ltd, 2018) Singh, V. P.; Badiger, N. M.; Korkut, Turgay
    Mass attenuation coefficients and exposure buildup factors (EBF) for some ceramic hosts such as Hollandite (BaAl2Ti6O16), Perovskite (CaTiO3), Zirconolite (CaZrTi2O7), Apatite (Pb-10 (VO4)(4.8)(PO4)(1.2)I-2), and Zircon (ZrSiO4) for high level radioactive waste have been computed in the present paper. The mass attenuation coefficients for the Apatite were found to be the highest. The EBF for the Apatite were found the smallest in low-to-intermediate energy (<3 MeV). Neutron total macroscopic cross sections for the ceramic hosts were calculated for 2, 4.5 and 14.1 MeV using Geant4. Zircon for neutron low-energy (2 MeV) and Hollandite for high-energy (14.1 MeV) were found superior shielding materials. This study could be useful for radioactive waste management, handling, transportation, dose evaluation and other shielding requirements. (C) 2014 Elsevier Ltd. All rights reserved..
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    Öğe
    Gamma-ray and neutron shielding efficiency of Pb-free gadolinium-based glasses
    (Springer Singapore Pte Ltd, 2016) Singh, V. P.; Badiger, N. M.; Kothan, S.; Kaewjaeng, S.; Korkut, T.; Kim, H. J.; Kaewkhao, J.
    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) B2O3-10SiO(2)-10CaO-xGd(2)O(3) (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.