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    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
    New high temperature resistant heavy concretes for fast neutron and gamma radiation shielding
    (Walter De Gruyter Gmbh, 2019) Aygun, Bunyamin; Sakar, Erdem; Korkut, Turgay; Sayyed, Mohammed Ibrahim; Karabulut, Abdillhalik
    In the present work, we developed three new high temperature resistant heavy concretes as novel radiation shielding materials. For this purpose, chrome ore (FeCr2O4), hematite (Fe2O3), titanium oxide (TiO2), aluminum oxide (Al2O3), hematite [FeO(OH) nH(2)O], siderite (FeCO3), barite (BaSO4), nickel oxide (NiO) materials and alumina cement were used. Mass combination ratios of components and total macroscopic cross sections (scattering, absorption, capture, fission) of the samples were calculated by using GEANT4 code. The resistances of the prepared samples were evaluated in terms of compression strength after exposure at the 1000 degrees C temperature. Neutron equivalent dose rate measurements were carried out by using 4.5 MeV Am-241-Be neutron source and BF3 detector. All results were compared with normal weight concrete and paraffin. The results of neutron dose indicate that neutron absorption ability of the new heavy concretes is higher than normal weight concrete and paraffin. In addition to neutron measurements, different gamma-ray shielding parameters such as mass attenuation coefficient (MAC), effective atom numbers (Z(eff)), half value layer (HVL) and mean free path (MFP) have been calculated using WinXCOM software in order to investigate the effectiveness of using the prepared concretes as a radiation shielding materials. Gamma-ray results were compared with concretes and Pb-based glass.