Yazar "Gunay, M." seçeneğine göre listele

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    Calculations of neutron-induced production cross-sections of 180,182,183,184,186W up to 20 MeV
    (Pergamon-Elsevier Science Ltd, 2009) Sarer, B.; Aydin, A.; Gunay, M.; Korkmaz, M. E.; Tel, E.
    Neutron-induced cross-sections for the stable isotopes W-180,W-182,W-183,W-184,W-186 in the energy region up to 20 MeV have been calculated. Calculations were made with the codes CEM03.01 and ALICE/ASH, using the following models: the Dubna version of the intranuclear cascade model for the cascade stage of interaction; the hybrid, the geometry dependent hybrid and the exciton model for the pre-equilibrium component; the Hauser-Feshbach and the Weisskopf-Ewing statistical models for the equilibrium component. Effects of some important model parameters such as level density parameter and pairing correction were investigated. Calculated cross-sections were compared with available experimental data in the literature and with ENDF/B-VI T = 300 K and JENDL-3.3 T = 300 K evaluated data libraries. (C) 2008 Elsevier Ltd. All rights reserved.
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    Contributions of each isotope in some fluids on neutronic performance in a fusion-fission hybrid reactor: a Monte Carlo method
    (Indian Assoc Cultivation Science, 2014) Gunay, M.; Sarer, B.; Kasap, H.
    In the present investigation, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa ferritic steel structural material and 99-95 % Li20Sn80-1-5 % SFG-Pu, 99-95 % Li20Sn80-1-5 % SFG-PuF4, 99-95 % Li20Sn80-1-5 % SFG-PuO2 the molten salt-heavy metal mixtures, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium zone with the width of 3 cm was used for the neutron multiplicity between liquid first wall and blanket. The contributions of each isotope in fluids on the nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional analyses were performed by using Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.
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    The Effect on Neutronic Calculations of Certain Alternative Fuels in a Boiling Water Reactor by Using MCNPX Monte Carlo Method
    (Polish Acad Sciences Inst Physics, 2015) Gunay, M.; Espinoza, V. H. Sanchez; Travleev, A.
    We were studying on the modeling of boiling water reactor fuel assemblies at pin-by-pin level by using Monte Carlo method. The designed boiling water reactor system is cylinder, and the radius of the cylinder is 228 cm. The total active core height is 315.79 cm. The reactor core was divided into the square lattice 7 x 7 type with a constant pitch of 30 cm. The core was surrounded with the reflector. The reflector was surrounded by SS316LN ferritic steel with width of 3 cm. The mixtures 0.21-% PuF4 and PuO2 were used as fuel. In this study, the effect on the neutronic calculations of PuF4 and PuO2 fuels was investigated in the designed boiling water reactor system. There were calculated k(eff), heat deposition and the fission energy in the designed boiling water reactor system. The three-dimensional (3D) modelling of the reactor core and fuel assembly into the designed boiling water reactor system was performed by using MCNPX-2.7.0 Monte Carlo method and the ENDF/B nuclear data library.
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    Fluids in a Hybrid System by Using MCNPX Monte Carlo Radiation Transport Code
    (Polish Acad Sciences Inst Physics, 2015) Gunay, M.
    In this study, the effect of the radiation damage of spent fuel-grade plutonium content was investigated in the structural material of a designed fusion-fission hybrid reactor system. In this study, the molten salt-heavy metal mixtures 99-95% Li20Sn80-1-5% SFG-Pu, 99-95% Li20Sn80-1-5% SFG-PuF4, and 99-95% Li20Sn80-1-5% SFG-PuO2 were used as fluids. The fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. Proton, deuterium, tritium, He-3 and He-4 gas production rates are the parameters of radiation damage. In this study, damage to the total structural material and each 1.0 cm thickness thereof was measured as a function of radiation energy, using the selected fluid rates, for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII. 0 nuclear data library.
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    Investigation of (n,?) reaction in hybrid reactor zones
    (Carl Hanser Verlag, 2014) Gunay, M.
    In this study, the fluids were composed with increased mole fractions of a mixture of molten salt: heavy metals 99-95 % Li20Sn80-1 - 5 % SFG-Pu, 99 - 95 % Li20Sn80-1 - 5 % SFG-PuF4, 99 - 95 % Li20Sn80-1 - 5 % SFG-PuO2. In this study, the effect on conversion of each isotope (Pu238-242) in spent fuel grade plutonium by (n,gamma) reactions was investigated in liquid first wall, blanket and shield zones of the designed hybrid reactor system. Beryllium (Be) is the neutron multiplier by (n,2n) reactions. The Be zone used was 3 cm thick. 9Cr2WVT, a ferritic steel, is used as structural material. Three-dimensional nucleonic calculations were performed by using the most recent versions of the MCNPX-2.7.0 Monte Carlo code and the nuclear data library ENDF/B-VII.0.
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    Investigation of (n,?) reactions in fissionable fluids in a hybrid reactor system
    (Carl Hanser Verlag, 2014) Gunay, M.
    In this study, the effect of (n,gamma) reactions in spent fuel-grade (SFG)-PuO2, UO2, NpO2 and UCO contents was investigated in a designed hybrid reactor system. In this system, the molten salt-heavy metal mixtures 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2-10% UO2, 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2-10 % NpO2, 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2 10 % UCO were used as fluids. Beryllium (Be) is used as neutron multiplier by (n,2n) reactions. Thence, a Be zone with a width of 3 cm was used in order to contribute on fissile fuel breeding between the liquid first wall and blanket. (n,gamma) reactions were calculated in the liquid first wall, blanket and shield zones, which SFG-PuO2, UO2, NpO2 and UCO contents. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code and nuclear data library ENDF/B-VII.0.
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    Investigation of neutronic effects in structural material of a hybrid reactor by using the MCNPX Monte Carlo transport code
    (Carl Hanser Verlag, 2013) Gunay, M.
    In this study, 98-90% Li2BeF4-2-10 % ThF4, 98-90% Li2BeF4-2-10 % UF4, 98-90% Li2BeF4-2-10 % UO2 and 100% Li2BeF4 molten salt-heavy metal was used as fluid. The fluids were used in the liquid first wall, liquid second wall and shield zones of the designed hybrid reactor system. A steel wall of 4 cm thickness is used as structural material. Proton, deuterium, tritium, He-3 and He-4 gas production rates are the parameters of radiation damage. In the study, the effect of liquid second wall thicknesses (20 cm, 30 cm, 40 cm, 50 cm) on the neutron flux distribution and the parameters of radiation damage according to neutron energy spectrum in the structural material were investigated for the selected fluids. A three-dimensional analysis was done by using the most recent version of the MCNPX-2.7.0 Monte Carlo code and the nuclear data library ENDF/B-VII.
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    The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system
    (Carl Hanser Verlag, 2015) Gunay, M.
    In the present investigation, a hybrid reactor system was designed by using 100% Flibe, 90% Flibe-10% ThF4, 90 % Flibe-10% UF4 fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.
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    Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method
    (Carl Hanser Verlag, 2015) Gunay, M.
    In this study, salt-heavy metal mixtures consisting of 93-85% Li20Sn80+5% SFG-PuO2 and 2-10% UO2, 93-85% Li20Sn80+5% SFG-PuO2 and 2-10% NpO2, and 93-85% Li20Sn80+5% SFG-PuO2 and 2-10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion-fission hybrid reactor system. A beryllium (Be) zone with a width of 3 cm was used for neutron multiplicity between the liquid first wall and the blanket. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The contributions of each isotope in the fluids to the nuclear parameters, such as tritium breeding ratio (TBR), energy multiplication factor (M), and heat deposition rate, of the fusion-fission hybrid reactor were calculated in the liquid first wall, blanket, and shield zones. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.
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    The production of 238-242Pu(n,?)239-243Pu fissionable fluids in a fusion-fission hybrid reactor
    (Carl Hanser Verlag, 2014) Gunay, M.
    In this study, the effect of spent fuel grade plutonium content on Pu259-243 was investigated in a designed hybrid reactor system. In this system, the fluids were composed of a molten salt, heavy metal mixture with increased mole fractions 99-95% Li20Sn80-1-5% SFG-Pu, 99-95% Li20Sn80-1-5% SFG-PuF4, 99-95% Li20Sn80-1-5% SFG-PuO2. Beryllium (Be) is a neutron multiplier by (n,2n) reactions. Thence, a Be zone of 3 cm thickness was used in order to contribute to fissile fuel breeding between the liquid first wall and a 9Cr2WVTa ferritic steel blanket which is used as structural material. The production of Pu238-242(n,gamma)Pu239-243 was calculated in liquid first wall, blanket and shielding zones. Three-dimensional nucleonic calculations were performed by using the most recent version MCNPX-2.7.0 Monte Carlo code and nuclear data library ENDF/B-VII.0.