Yazar "Korkut, Turgay" seçeneğine göre listele

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    Cross Section Calculations of Neutron Induced Reactions on 124,126,128,134,136Xe
    (Springer, 2015) Kara, Ayhan; Yigit, Mustafa; Korkut, Turgay; Tel, Eyyup
    In the nuclear energy applications Xenon is a desirable material because of its high molecular weight and inert nature. The cross section calculations of Xe-124(n,2n)Xe-123, Xe-126(n,2n)Xe-125, Xe-128(n,2n)Xe-127, Xe-134(n,2n)Xe-133 and Xe-136(n,2n)Xe-135 reactions were performed by ALICE/ASH (with Weisskopf-Ewing equilibrium model), TALYS 1.6 (with two component exciton model suggested by Kalbach) and EMPIRE 3.2 Malta (with exciton model recommended by Clinev and Ribansky) nuclear reaction codes. In addition, semi-empirical cross section formula of Tel et al. was used to obtain cross sections at 14-15 MeV energy. Obtained results were compared to available experimental data by EXFOR database and TENDL-2013 data.
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    Detailed Monte Carlo simulations on secondary radiation emissions of polymer based neutron shielding materials
    (Iop Publishing Ltd, 2024) Korkut, Hatun; Korkut, Turgay
    Since radiation safety is a sensitive issue in terms of health, this situation poses a significant problem. In order to raise awareness among researchers while preparing future studies on this problem, in this study, the interactions between fast neutrons and 10 different polymer-based shielding materials determined from the literature were modeled with the GEANT4 Monte Carlo code. The fast neutron shielding performances of the samples and the emitted secondary radiation and radioactivity have been reported. When evaluated number and mean energies of created particles, in order to solve this problem, the sum of the risk factors obtained by multiplying these values for each sample can provide us with the opportunity to make a more accurate evaluation. When these values are examined, sample P9 is at the top of the risk ranking with a value of 1823472, while sample P6 is at the bottom with a value of 146246. As a result, it has been revealed that these samples, which have good fast neutron shielding properties, also produce high levels of radiation and radioactive nuclei.
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    Determination and Fabrication of New Shield Super Alloys Materials for Nuclear Reactor Safety by Experiments and Cern-Fluka Monte Carlo Simulation Code, Geant4 and WinXCom
    (Iop Publishing Ltd, 2016) Aygun, Bunyamin; Korkut, Turgay; Karabulut, Abdulhalik
    Despite the possibility of depletion of fossil fuels increasing energy needs the use of radiation tends to increase. Recently the security-focused debate about planned nuclear power plants still continues. The objective of this thesis is to prevent the radiation spread from nuclear reactors into the environment. In order to do this, we produced higher performanced of new shielding materials which are high radiation holders in reactors operation. Some additives used in new shielding materials; some of iron (Fe), rhenium (Re), nickel (Ni), chromium (Cr), boron (B), copper (Cu), tungsten (W), tantalum (Ta), boron carbide (B4C). The results of this experiments indicated that these materials are good shields against gamma and neutrons. The powder metallurgy technique was used to produce new shielding materials. CERN - FLUKA Geant4 Monte Carlo simulation code and WinXCom were used for determination of the percentages of high temperature resistant and high-level fast neutron and gamma shielding materials participated components. Super alloys was produced and then the experimental fast neutron dose equivalent measurements and gamma radiation absorpsion of the new shielding materials were carried out. The produced products to be used safely reactors not only in nuclear medicine, in the treatment room, for the storage of nuclear waste, nuclear research laboratories, against cosmic radiation in space vehicles and has the qualities.
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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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    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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    Improving neutron and gamma flexible shielding by adding medium-heavy metal powder to epoxy based composite materials
    (Pergamon-Elsevier Science Ltd, 2019) Canel, Aysegul; Korkut, Hatun; Korkut, Turgay
    Epoxy resin is cheap, flexible, durable and abundant material. Within this framework, it is a very useful option in a comprehensive application field. In this paper, neutron shields based on epoxy based composite materials were studied by adding different percentages of additive materials as Fe (Iron), Bi (Bismuth), Ta (Tantalum) and WC (Tungsten Carbide). GEANT4 and FLUKA simulation codes were applied simulating and comparing the interactions of 4.5 MeV Am-241-Be neutrons and 1.25 MeV Co-60 source energetic gamma rays with the different specific materials. The best shielding performance results were obtained with the Tantalum-Epoxy sample but all the studied samples demonstrate a good quality as neutron and gamma shielding, when compared to previous results found in literature. Implications of neutron and gamma bombardment were also discussed in the view of a large usage fields.
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    Interaction of Radiation with Matter and Related Topics
    (Hindawi Ltd, 2014) Kaewkhao, Jakrapong; Djamal, Mitra; Korkut, Turgay
    [No abstract available]
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    Investigation of high-temperature-resistant rhenium-boron neutron shields by experimental studies and Monte Carlo simulations
    (Springer Singapore Pte Ltd, 2018) Korkut, Turgay; Korkut, Hatun; Aygun, Bunyamin; Bayram, Ozkan; Karabulut, Abdulhalik
    In this study, novel rhenium-boron neutron-shielding high-temperature-resistant materials were designed. The considered samples, Re60-B40, Re58-B42, Re50-B50, and Re40-B60, with different concentrations of rhenium and boron were investigated to elucidate their neutron-shielding performances, and compare them with well-known neutron-shielding materials such as the 316LN quality nuclear steel. In addition to the experimental studies, Monte Carlo simulations were performed using the FLUKA and GEANT4 codes, where 4.5-MeV neutrons emitted by a Am-241-Be source were employed. Experimental equivalent dose rates, simulated track lengths, energy balances, and neutron mass absorption cross sections were discussed in detail.
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    Monte Carlo Simulations of 17.9-22.3 MeV Energetic Proton Irradiation Effects on bcc-Zirconium Fusionic Materials
    (Springer, 2016) Korkut, Hatun; Korkut, Turgay; Kara, Ayhan; Yigit, Mustafa; Tel, Eyyup
    Advanced fusion structural materials (FSMs) have impact role in terms of efficiency of nuclear energy production. Besides engineering and design of fusion reactors, radiation durability of FSMs is another valuable issue that cannot be ignored. 17.9-22.3 MeV proton irradiation of bcc-Zirconium Fusion Structural Material was evaluated by using Monte Carlo based simulation tools. Total binary reaction cross sections were respectively calculated as 1167.6 and 1273.92 mb for 17.3- and 22.3 MeV proton energies via TALYS-1.6 version. Additionally, residual production cross sections and total particle production cross sections were obtained and analyzed by the TALYS code. Radiation damage parameters as Displacement Per Atom (DPA) and Stopping Power (SP) were studied by SRIM-2013 version. FLUKA 2011.1 used for only DPA calculations and making a complete comparison with the other calculation results. SP and Number of Secondaries were found by using GEANT4.10.p.04 version simulations. Natural Zr(p,x) reactions were studied in the given energy values in the plane of reaction probability and radiation damage calculations.
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    Monte Carlo simulations of proton irradiation effects on PTFE, KEVLAR, EFTE, DYNEEMA, and NOMEX polymeric materials used in astronaut space suits
    (Taylor & Francis Ltd, 2023) Korkut, Turgay; Korkut, Hatun
    Astronauts suffer from natural space radiation as high-energetic protons, heavy ions, and secondary particles produced in collisions. Galactic cosmic rays and solar particle events are the basic parts of space radiation spectra. Wears and accessories designs for use by astronauts aim to minimize these deleterious effects of this environment. Polymeric materials are preferred in astronaut suits because they are lightweight, inexpensive, and durable. Teflon, KEVLAR, ethylene tetrafluoroethylene, DYNEEMA, and NOMEX polymeric astronaut wear materials were exposed to high energetic proton irradiations by the use of Monte Carlo simulation tools SRIM-2013 and FLUKA 2011.1. Proton energies are applied as 1, 2, and 3 GeV known as in order of space radiation magnitude. Besides, displacement per atom calculations were done and results were discussed in the plane of structural changes in the given polymeric materials. After interacting with protons with 1, 2, and 3 GeV energies, the material with the lowest Displacement per atom value among the five studied materials is DYNEEMA with values of 1.01E - 25, 9.96E - 26, and 1.00E - 25, respectively. Again, among the materials studied for these three proton energies, DYNEEMA has the highest electronic stopping power values are, respectively, 2.11E - 03, 2.10E - 03, and 2.31E - 03. DYNEEMA has the highest nuclear stopping power values as 2.23E - 07, 1.53E - 07, and 4.27E - 07, respectively.
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    Neutron Radiation Tests about FeCr Slag and Natural Zeolite Loaded Brick Samples
    (Hindawi Ltd, 2014) Cay, Vedat Veli; Sutcu, Mucahit; Gencel, Osman; Korkut, Turgay
    Neutron shielding performances of new brick samples are investigated. Brick samples including 10, 20, and 30 percentages of ferrochromium slag (FeCr waste) and natural zeolite are prepared and mechanical properties are obtained. Total macroscopic cross sections are calculated by using results of 4.5MeV neutron transmission experiments. As a result, neutron shielding capacity of brick samples increases with increasing FeCr slag and natural zeolite percentages. This information could be useful in the area of neutron shielding.
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    Neutron shielding qualities and gamma ray buildup factors of concretes containing limonite ore
    (Elsevier Science Sa, 2015) Oto, Berna; Yildiz, Nergiz; Korkut, Turgay; Kavaz, Esra
    Neutron dose transmissions for fast neutrons produced by 5.486 MeV alpha particles on beryllium are measured in concrete samples with and without limonite ore to investigate their neutron shielding qualities. Using measured values, macroscopic removal cross-sections (Sigma(R), cm(-1)) have been determined experimentally and also Sigma(R)values have been calculated theoretically using the elemental composition of the concrete mixes. The best neutron shielding property of concrete sample containing 100% limonite ore (FCL: fine and coarse limonite) was found. Additionally, energy absorption buildup factor (EABF) and exposure buildup factor (EBF) of concrete sample were calculated using the five-parameter Geometric Progression (G-P) approximation in the energy range of 0.015-15 MeV for penetration depths up to 40 mean free path (mfp). The incident photon energy and penetration depth dependence of buildup factors were examined. Finally, we observed that concrete samples have maximum values of buildup factors in the intermediate energy region around 0.1-0.3 MeV. FCL has the minimum values of both of the buildup factors. FCL has the excellent gamma shielding properties compared to the concrete samples. (C) 2015 Elsevier B.V. All rights reserved.
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    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.
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    Production and Neutron Irradiation Tests on a New Epoxy/Molybdenum Composite
    (Taylor & Francis Ltd, 2015) Aygun, Buenyamin; Korkut, Turgay; Karabulut, Abdulhalik; Gencel, Osman; Karabulut, Abdulkerim
    Molybdenum powder was added in an epoxy matrix to increase the neutron-shielding capacity of pure epoxy. FLUKA and GEANT4 Monte Carlo programs were used to design new neutron-shielding material. After finding the epoxy/Mo ratio for the best shielding capacity, neutron equivalent dose rate measurements were performed. Results were compared to commonly used neutron-shielding materials such as paraffin, steel, concrete, and B4C. The produced new sample has a high neutron shielding capacity and it can be used for neutron protection purposes.
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    Simulations on the performances of neutron shielding glass materials and secondary radiation risks
    (Pergamon-Elsevier Science Ltd, 2024) Korkut, Hatun; Korkut, Turgay
    Glasses with neutron shielding properties are extremely important for nuclear industries. Neutron interactions with ten different neutron shielding glass types were simulated by using GEANT4 Monte Carlo simulation toolkit. Simulation results were evaluated in terms of shielding performances, deposited energy properties and produced secondary radiation. Important evaluations were made in view of radiation safety, especially on secondary particles and radioactivity resulting from interactions between neutrons and glass samples. All simulated glass samples were found to emit risky secondary particles and radioactive nuclei after interacting with fast neutrons.
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    Study of neutron attenuation properties of super alloys with added rhenium
    (Springer, 2015) Korkut, Turgay; Aygun, Bunyamin; Bayram, Ozkan; Karabulut, Abdulhalik
    In this paper, the new generation of rhenium enriched super alloys with enhanced neutron attenuation properties was investigated. These materials provided improved neutron shielding capabilities compared to other shielding materials such as water, concrete, etc. The GEANT4 code was used to simulate transport of Am-Be source neutrons in new shielding materials. Neutron equivalent dose rate measurements were carried out to obtain the dose rate reduction. It was demonstrated that new rhenium enriched alloys are feasible for neutron shielding applications.