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    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.
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    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.
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    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.
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    Monte Carlo studies in accelerator-driven systems for transmutation of high-level nuclear waste
    (Pergamon-Elsevier Science Ltd, 2008) Sarer, Basar; Korkmaz, M. Emin; Guenay, Mehtap; Aydin, Abdullah
    A spallation neutron source was modeled using a high energy proton accelerator for transmutation of Pu-239, minor actinides Np-237, Am-241 and long-lived fission products Tc-99, I-129, which are created from the operation of nuclear power reactors for the production of electricity. The acceleration driven system (ADS) is composed of a natural lead target, beam window, subcritical core, reflector, and structural material. The neutrons are produced by the spallation reaction of protons from a high intensity linear accelerator in the spallation target, and the fission reaction in the core. It is used a hexagonal lattice for the waste and fuel assemblies. The system is driven by a 1 GeV, 10 mA proton beam incident on a natural lead cylindrical target. The protons were uniformly distributed across the beam. The core is a cylindrical assembly. The main vessel is surrounded by a reflector made of graphite. The axes of the proton beam and the target are concentric with the main vessel axis. The structural walls and the beam window are made of the same material, stainless steel, HT9. We investigated the following neutronics parameters: spallation neutron and proton yields, spatial and energy distribution of the spallation neutrons, and protons, heat deposition, and the production rates of hydrogen and helium, transmutation rate of minor actinides and fission products. In the calculations, the Monte Carlo code MCNPX, which is a combination of LAHET and MCNP, was used. To transport a wide variety of particles, The Los Alamos High Energy Transport Code (LA HET) was used. (C) 2007 Elsevier Ltd. All rights reserved.
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    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.
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    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.