Yazar "Gunay, Mehtap" seçeneğine göre listele

Listeleniyor 1 - 13 / 13
  • Küçük Resim Yok
    Öğe
    Assessment of the neutronic performance of some alternative fluids in a fusion-fission hybrid reactor by using Monte Carlo method
    (Pergamon-Elsevier Science Ltd, 2013) Gunay, Mehtap
    In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVFa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80-1-5% SFG-Pu, 99-95% Li20Sn80-1-5% SFG-PuF4, or 99-95% Li20Sn80-1-5% SFG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as neutron flux, tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fissile fuel breeding in liquid first wall, blanket and shield zones and investigates effects of spent fuel grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.O. Published by Elsevier Ltd.
  • Küçük Resim Yok
    Öğe
    The calculation of neutron flux using Monte Carlo method
    (E D P Sciences, 2017) Gunay, Mehtap; Bardakci, Hilal
    In this study, a hybrid reactor system was designed by using 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2 fluids, ENDF/B-VII. 0 evaluated nuclear data library and 9Cr2WVTa structural material. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The neutron flux was calculated according to the mixture components, radial, energy spectrum in the designed hybrid reactor system for the selected fluids, library and structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.
  • Küçük Resim Yok
    Öğe
    COMPARISONS OF THE CALCULATIONS USING DIFFERENT CODES IMPLEMENTED IN MCNPX MONTE CARLO TRANSPORT CODE FOR ACCELERATOR DRIVEN SYSTEM TARGET
    (Amer Nuclear Soc, 2012) Sarer, Basar; Sahin, Sumer; Gunay, Mehtap; Celik, Yurdunaz
    The MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation-fission models and CEM2k cascade-exciton model. The study analyzes the main quantities determining ADS performance such as neutron yield, neutron leakage spectra, and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models, cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. Target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. Target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 5.3 cm and an inner radius 5.0 cm. The maximum of the neutron flux in the target is observed on the axis similar to 10 cm below the beam window, where the maximum difference between 7 different models is similar to 15 %. The total neutron leakage out of the of the target calculated with the Bertini/ABLA is 1.83x10(17) n/s, and is about 14 % higher than the value calculated by the INCL4/Dresner (1.60x10(17) n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20 %, 2.3 %, 77.6 % of the total leakage, respectively, whereas, they become 18.6 %, 2.3 %, 79.4 % with INCL4/Dresner combination.
  • Küçük Resim Yok
    Öğe
    Contribution of each isotope in structral material on radiation damage in a hybrid reactor
    (E D P Sciences, 2016) Gunay, Mehtap
    In this study, the fluids were used in the liquid first-wall, blanket and shield zones of the designed hybrid reactor system. 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. In this study, the effect on the radiation damage of spent fuel-grade (SFG)-PuO2, UO2, NpO2 and UCO contents was investigated in the structural material of a designed fusion-fission hybrid reactor system. In the designed hybrid reactor system were investigated the effect on the radiation damage of the selected fluid according to each isotopes of structural material in the structural material 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.
  • Küçük Resim Yok
    Öğe
    The effect on radiation damage of structural material in a hybrid system by using a Monte Carlo radiation transport code
    (Pergamon-Elsevier Science Ltd, 2014) Gunay, Mehtap; Sarer, Basar; Kasap, Hizir
    In this study, the molten salt-heavy metal mixtures 99-95% Li20Sn80-1-5% SFG-Pu, 99-95% Li20Sn80-1-5% SFG-PuF4, 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. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The parameters of radiation damage are proton, deuterium, tritium, He-3 and He-4 gas production rates. In this study, the effects of the selected fluid on the radiation damage, in terms of individual as well as total isotopes in the structural material, were investigated for 30 full power years (FPYs). Three-dimensional analyses were performed using the most recent version of the MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library. Published by Elsevier Ltd.
  • Küçük Resim Yok
    Öğe
    Evaluation of integral quantities in an accelerator driven system using different nuclear models implemented in the MCNPX Monte Carlo transport code
    (Pergamon-Elsevier Science Ltd, 2013) Sarer, Basar; Sahin, Sumer; Celik, Yurdunaz; Gunay, Mehtap
    The MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation-fission models and CEM2k cascade-exciton model. This study analyzes the main quantities determining ADS performance, such as neutron yield, neutron leakage spectra, heating and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. The target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. The target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 53 cm and an inner radius of 5.0 cm. The maximum value of the neutron flux in the target is observed on the axis similar to 10 cm below the beam window, where the maximum difference between 7 different models is similar to 15%. The total neutron leakage of the target calculated with the Bertini/ABLA is 1.83 x 10(17) n/s, and is about 14% higher than the value calculated by the INCL4/Dresner (1.60 x 10(17) n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20%, 2.3%, 77.6% of the total leakage, respectively, whereas, the calculated fractions are 18.6%, 2.3%, 79.4%, respectively, with INCL4/Dresner combination. The largest heat deposition density by considering all particles in the beam window calculated with CEM2k model is 104 W/cm(3)/mA, which is 9.0% greater than the lowest value predicted with INCL4/Dresner model (95.4 W/cm(3)/mA). The maximum average heat deposition density for all particles in the target is calculated as 6.87 W/cm(3)/mA with INCL4/ABLA. (C) 2013 Elsevier Ltd. All rights reserved.
  • Küçük Resim Yok
    Öğe
    Investigation of radiation damage in structural material of APEX reactor by using Monte Carlo method
    (Pergamon-Elsevier Science Ltd, 2013) Gunay, Mehtap
    In this study, ThF4, UF4 and UO2 heavy metals were used with ratios of 2% and 10% while Flibe (Li2BeF4) molten salt mixture and 100% Flibe were used as fluids in the liquid first wall, liquid second wall and shield zones of the APEX. The steel wall that is used as a structural material is 4 cm in thickness and calculations for each 0.5 cm thick zone of the steel wall were performed. In this study, the total changes caused by radiation damage in the steel used as a structural material for 20 cm, 30 cm, 40 cm and 50 cm thicknesses of liquid second wall, for each 0.5 cm thick zones of the steel wall and for the selected fluid rates were investigated for 30 full power years (FPY). The neutron wall load is assumed to be 10 MW/m(2). Three-dimensional nucleonic calculations were performed using MCNPX-2.5.0 Monte Carlo code and ENDF/B-VI nuclear data library. Published by Elsevier Ltd.
  • Küçük Resim Yok
    Öğe
    Monte Carlo calculations of the nuclear effects of certain fluids in a hybrid reactor
    (Pergamon-Elsevier Science Ltd, 2016) Gunay, Mehtap
    This study analyzes the effects of certain heavy-metal-salt fluids on nuclear parameters in a fusion-fission hybrid reactor. Calculated parameters include the tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate, and fissile fuel breeding in the reactor's liquid first wall, blanket, and shield zones; gas production rates in the structural material of the reactor were calculated, as well. The fluid mixtures consisted of 93-85% Li20Sn50 + 5% SFG-PuO2 and 2-10% UO2, 93-85% Li20Sn80 + 5% SFG-PuO2 and 2-10% NpO2, and 93-85% Li20Sn50 + 5% SFG-PuO2 and 2-10% UCO. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion fission hybrid reactor system. A 3 cm wide beryllium (Be) zone was used for neutron multiplier between the liquid first wall and the blanket. The structural material used was 9Cr2WVTa ferritic steel, measuring 4 cm in width. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and the ENDF/B-VILO nuclear data library. (C) 2015 Elsevier Ltd. All rights reserved.
  • Küçük Resim Yok
    Öğe
    Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code
    (Carl Hanser Verlag, 2014) Gunay, Mehtap
    In this study, 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. 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. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. 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.
  • Küçük Resim Yok
    Öğe
    Neutronic investigation of the application of certain plutonium-mixed fluids in a fusion-fission hybrid reactor
    (Pergamon-Elsevier Science Ltd, 2014) Gunay, Mehtap; Kasap, Hizir
    In this study, the fluids that were investigated were contained increased mole fractions of the mixtures molten salt 99-95% Li20Sn80 and, the heavy metals 1-5% SFG-Pu, SFG-PuF4, SFG-PuO2. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplicity between liquid first wall and blanket. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. The nuclear parameters of a fusion-fission hybrid reactor such as neutron flux, heating, fission reaction rate were investigated according to the mixture components, radial energy spectrum in the designed system. In this study, the effect of spent fuel-grade Pu content in the designed system on these nuclear parameters were calculated by using the three-dimensional Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0. Crown Copyright (C) 2013 Published by Elsevier Ltd. All rights reserved.
  • Küçük Resim Yok
    Öğe
    Three-dimensional Monte Carlo calculation of gas production in structural material of APEX reactor for some evaluated data files
    (Pergamon-Elsevier Science Ltd, 2013) Gunay, Mehtap; Sarer, Basar; Hancerliogullari, Aybaba
    Proton and He-4 gas production rates in the structural material of a fusion-fission hybrid reactor were calculated with the three-dimensional Monte Carlo method by the MCNPX-2.5.0 code. We examined these reaction rates with five nuclear data libraries: ENDF/B-VII.0 T = 300 K, JEFF-3.1 T = 300 K, JENDL-4.0 T = 300 K, ROSFOND T = 300 K and CENDL-3.1 T = 300 K. The production from each isotope of structural material made of ferritic steel was calculated. The neutron flux load is assumed to be 10 MW/m(2). (C) 2013 Elsevier Ltd. All rights reserved.
  • Küçük Resim Yok
    Öğe
    Three-dimensional Monte Carlo calculation of some nuclear parameters
    (E D P Sciences, 2017) Gunay, Mehtap; Seker, Gokmen
    In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Threedimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.
  • Küçük Resim Yok
    Öğe
    Three-dimensional neutronic calculations for a fusion breeder APEX reactor using some libraries
    (Pergamon-Elsevier Science Ltd, 2011) Gunay, Mehtap; Sarer, Basar; Celik, Yurdunaz
    The effects of evaluated nuclear data files on neutronics characteristics of a fusion-fission hybrid reactor have been analyzed; three-dimensional calculations have been made using the MCNP4C Monte Carlo Code for ENDF/B-VII T= 300K. JEFF-3.0 T= 300K, and CENDL-2 T= 300 K evaluated nuclear data files. The nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, fissile fuel breeding and nuclear heating in a first wall, blanket and shield have been investigated for the mixture components of 90% Flibe (Li(2)BeF(4)) and 10% UF(4) for a blanket layer thickness of 50 cm. The contributions of each isotope of Flibe ((6)Li, (7)Li, (19)F, (9)Be) and UF(4) ((235)U, (238)U) to the integrated parameter values were calculated. The neutron wall load is assumed to be 10 MW/m(2). (C) 2011 Elsevier Ltd. All rights reserved.