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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.