Yazar "Sahin, Haci Mehmet" seçeneğine göre listele

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    Development of an air velocity and flow measurement system by using a novel circular disc
    (Elsevier Sci Ltd, 2021) Sahin, Haci Mehmet; Soysal, M. Saim
    In this paper, development of a sensor using a circular disc for air velocity measurement based on the drag force equation is proposed. The air velocity measurement sensor is basically designed with a load cell in order to determine the drag force. The circular disc is used for creating a drag force, and by using load cell, the drag force that acts on the circular disc is measured. As the circular disc's drag coefficient can be considered constant at Reynolds numbers between 10(3) and 10(6), it can be possible to obtain the explicit equation of drag force. The remaining components of drag force equation are obtained by measurement. The proposed air velocity measurement sensor is characterized by wind tunnel measurements. All measurements were performed in an ISO/IEC 17025 accredited calibration laboratory - the Wind Tunnel of Turkish State Meteorological Service. The characterization measurements were performed at air velocities between 1 m/s and 20 m/s. The correction factors were calculated and a calibration curve for the proposed air velocity measurement sensor was obtained. The calibration curve's linearity was higher than 0.99 and a comparison the results from a Micromanometer with Pitot-Tube shows that for the designed working range, the sensor has an acceptable performance for time-averaged air velocity measurements according to the requirements of the World Meteorological Organization.
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    Generation-IV reactors and nuclear hydrogen production
    (Pergamon-Elsevier Science Ltd, 2021) Sahin, Sumer; Sahin, Haci Mehmet
    There are dozens of hydrogen production methods and techniques from many sources such as fossil fuels, renewable energy sources and nuclear energy in the literature. Thermo-chemical methods are more efficient at higher temperatures to produce large quantities of hydrogen. In this study, a comparative overview of Generation VI nuclear reactor types for major hydrogen production methods have been researched in the literature and suggestions have been carried out. This research work is addressing that both electric power cycle and hydrogen production based on nuclear technologies need to be developed. Generation IV nuclear reactors can provide hydrogen for a worldwide hydrogen economy. Both thermo-chemical and electrolysis (hybrid) processes in hydrogen production have a promising future, especially when integrated with Generation IV nuclear power plants. Efficient heat transfer is required for both high temperature thermodynamic cycles and the high temperature steam electrolysis. Hence, highly efficient heat exchanger designs are one of the key technologies for that purpose. (c) 2020 Hydrogen Energy Publications LLC. Published by Elsevier Ltd. All rights reserved.
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    Neutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixture
    (Pergamon-Elsevier Science Ltd, 2024) Sahin, Haci Mehmet; Tunc, Guven; Karakoc, Alper
    In this study, a novel approach has been investigated by using a mixture of thorium and molten salts as a dualpurpose coolant and medium for the production of fissile fuel in a Fusion Fission Hybrid Reactor (FFHR) for the reference geometry of ITER. The study highlighted the broader benefits of thorium fuel cycling, safety features, and reduced radioactive minor actinides generation. The use of a thorium-melted salt coolant for fissile fuel production in a fusion-fission hybrid reactor represented a promising path towards efficient and sustainable nuclear energy, with potential benefits in terms of safety features and reduced generation of radioactive minor actinides. In this study, SS 316 LN-IG was selected as the first wall material for the reactor, and a molten salt fuel mixture of LiF-ThF4 was used as the coolant, taking into account the eutectic points of the material, the nominal fusion power in the FFHR for the Tokamak design concept is considered to be 500 MW. The nuclear code MCNP6 was used with the nuclear data libraries ENDF/B-VIII and CLAW-IV for the neutron calculations. The time evolution of the isotopes in the reactor was calculated with the interface code MCNPAS. The study results are evaluated in terms of tritium breeding ratio, energy multiplication factor, radiation damage, fissile fuel production and fuel burn-up value.The 4-year operation history of total TBR value is calculated and always above 1.05 and increases with time.Th initially decreased from 631.3 tonnes to 587.2 tonnes, while 233U production during this period was 9.1 tonnes. According to these results, the first wall replacement period was calculated as 3.9 years.
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    Neutronic study on the effect of first wall material thickness on tritium production and material damage in a fusion reactor
    (Springer Singapore Pte Ltd, 2022) Sahin, Haci Mehmet; Tunc, Guven; Karakoc, Alper; Omar, Melood Mohamad
    In this study, the effects of changing first wall materials and their thicknesses on a reactor were investigated to determine the displacement per atom (DPA) and gas production (helium and hydrogen) in the first wall, as well as the tritium breeding ratio (TBR) in the coolant and tritium breeding zones. Therefore, the modeling of the magnetic fusion reactor was determined based on the blanket parameters of the International Thermonuclear Experimental Reactor (ITER). Stainless steel (SS 316 LN-IG), Oxide Dispersion Strengthened Steel alloy (PM2000 ODS), and China low-activation martensitic steel (CLAM) were used as the first wall (FW) materials. Fluoride family molten salt materials (FLiBe, FLiNaBe, FLiPb) and lithium oxide (LiO2) were considered the coolant and tritium production material in the blanket, respectively. Neutron transport calculations were performed using the well-known 3D code MCNP5 using the continuous-energy Monte Carlo method. The built-in continuous energy nuclear and atomic data libraries along with the Evaluated Nuclear Data file (ENDF) system (ENDF/B-V and ENDF/B-VI) were used. Additionally, the activity cross-section data library CLAW-IV was used to evaluate both the DPA values and gas production of the first wall (FW) materials. An interface computer program written in the FORTRAN 90 language to evaluate the MCNP5 outputs was developed for the fusion reactor blanket. The results indicated that the best TBR value was obtained for the use of the FLiPb coolant, whereas depending on the thickness, the first wall replacement period in terms of radiation damage to all materials was between 6 and 11 years.
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    Study on the fusion reactor performance with different materials and nuclear waste actinides
    (Wiley, 2021) Sahin, Sumer; Sahin, Haci Mehmet; Sahiner, Huseyin; Tunc, Guven
    In this study, both pure fusion blanket and fusion-fission (hybrid) reactor blanket performance were investigated and discussed separately in two phases. In the first phase, a Monte Carlo radiation damage analysis has been performed for stainless steel (SS304, SS316, and oxide dispersion strengthened (ODS)), molybdenum, vanadium, and tungsten as the first wall (FW) materials, in combination with selected tritium breeders. The main technical parameters for fusion reactors, such as tritium breeding ratio, fusion energy multiplication factor (M), displacement per atom (DPA), and gas production (He, H) have been evaluated. All numerical calculations have been carried out in spherical geometry with MCNP6 code package using continuous energy cross-sections from the ENDF/B-VIII.0 library, except DPA calculations. Instead of the ENDF/B-VIII.0 library, the 30-group CLAW-IV library was employed for DPA calculations. Structural material selection for the FW respect to radiation damage limits and reactor performance for energy production and tritium has been concluded. Conventional thermal reactors, such as light water reactors andCanada Deuterium Uranium (CANDU) reactors are producing substantial quantities of transuranic elements, which represent serious nuisance and permanent hazard potential. On the other hand, they become fissionable material under high energetic fusion neutron irradiation and multiply the fusion energy. In the second phase, the investigations are extended to the incineration of minor actinides (MA) in the fusion-fission (hybrid) mode. The transmutation history of MA nuclear waste is included. MA are added into the first zone of the coolant in TRi-structural ISOtropic particle TRISO particles with a volume fraction of 6%. The transformation scenario for all MA by SS 304 steel FW is practically the same as with the ODS FW.