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Conference Object Citation - WoS: 8Citation - Scopus: 7UTILIZATION OF REACTOR GRADE PLUTONIUM AS ENERGY MULTIPLIER IN THE LIFE ENGINE(Amer Nuclear Soc, 2012) Sahin, Sumer; Sahin, Haci Mehmet; Acir, AdemThe accumulated reactor grade (RG)-plutonium as nuclear waste of conventional reactors is estimated to exceed 1700 tonnes. Laser Inertial Confinement Fusion Fission Energy (LIFE) engine is considered to incinerate RG-plutonium in stockpiles. Calculations have been conducted for a constant fusion driver power of 500 MWth in S-8-P-3 approximation using 238-neutron groups. RG-plutonium out of the nuclear waste of LWRs is used in form of fissile carbide fuel in TRISO particles with volume fractions of 2, 3, 4, 5 and 6 %, homogenously dispersed in the Flibe coolant. Respective tritium breeding ratio (TBR) values per incident fits ion neutron are calculated as TBR = 1.35, 1.52, 1.73, 2.02 and 2.47 at start-up. With the burn up of fissionable RG-Pu isotopes in the coolant, TBR decreases gradually. Similarly, blanket energy multiplications are calculated as M-0 = 3.8, 5.5, 7.7, 10.8 and 15.4 at start-up, respectively. Calculations have indicated prospects of achievability of very high burn up values (> 400 000 MD.D/MT).Conference Object Renewability and Sustainability Aspects of Nuclear Energy(Amer inst Physics, 2014) Sahin, SumerRenewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, U-233 fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO2/RG-PuO2) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG-PuO2 + 96 % ThO2; 6 % RG-PuO2 + 94 % ThO2; 10 % RG-PuO2 + 90 % ThO2; 20 % RG-PuO2 + 80 % ThO2; 30 % RG-PuO2 + 70 % ThO2, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as similar to 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of similar to 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG-PuO2 fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW. d/ton level. A laser fusion driver power of 500 MWth has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce similar to 160 kg U-233 per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of similar to 1.3.Article Citation - WoS: 7Citation - Scopus: 10Evaluation 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, MehtapThe 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.Editorial Editorials, "irec 2012, the International Renewable Energy Congress," Sousse, Tunisia (december 19-22, 2012)(Pergamon-elsevier Science Ltd, 2014) Sahin, Sumer[No Abstract Available]Editorial Citation - WoS: 1Citation - Scopus: 1Editor's Notes on Icenes'2013, 16th International Conference on Emerging Nuclear Energy Systems(Pergamon-elsevier Science Ltd, 2015) Sahin, Sumer[No Abstract Available]Article Citation - WoS: 6Assessment of Criticality and Burn Up Behavior of Candu Reactors With Nuclear Waste Trans Uranium Fuel(Pergamon-elsevier Science Ltd, 2012) Sahin, Sumer; Ahmed, Rizwan; Khan, Mohammad JavedLarge quantities of nuclear waste plutonium and minor actinides (MAs) have been accumulated in the civilian light water reactors (LWRs) and CANDU reactors. These trans uranium (TRU) elements are all fissionable, and thus can be considered as fissile fuel materials in form of mixed fuel with thorium or naturanium in the latter. CANDU fuel compacts made of tristructural-isotropic (TRISO) type pellets would withstand very high burn ups without fuel change. As carbide fuels allow higher fissile material density than oxide fuels, following fuel compositions have been selected for investigations: (1) 90% nat-UC + 10% TRUC, (2) 70% nat-UC + 30% TRUC and (3) 50% nat-UC + 50% TRUC. Higher TRUC charge leads to longer power plant operation periods without fuel change. The behavior of the criticality k(infinity) and the burn up values of the reactor have been pursued by full power operation for > similar to 12 years. For these selected fuel compositions, the reactor criticality starts by k(infinity) = 1.4443, 1.4872 and 1.5238, where corresponding reactor operation times and burn up values have been calculated as 2.8 years, 8 years and 12.5 years, and 62, 430 MW.D/MT, 176,000 and 280,000 MW.D/MT, with fuel consumption rates of similar to 16, 5.68 and 3.57 g/MW.D respectively. These high burn ups would reduce the nuclear waste mass per unit energy output drastically. The study has show clearly that TRU in form of TRISO fuel pellets will provide sufficient criticality as well as reasonable burn up for CANDU reactors in order to justify their consideration as alternative fuel. (c) 2012 Elsevier Ltd. All rights reserved.Conference Object Citation - WoS: 2Citation - Scopus: 3Influence of Void Fraction on Bwr Spent Fuel Direct Recycling Scenario(Pergamon-elsevier Science Ltd, 2015) Waris, Abdul; Su'ud, Zaki; Sahin, Hacz Mehmet; Kurt, Erol; Sahin, SumerPreliminary study on influence of changing void fraction (VF) on SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario for boiling water reactor (BWR) spent fuel direct recycling scheme has been carried out. Several VF values of BWR have been investigated to determine the criticality of reactor. The VF values range from 20% to 60%. The fraction of spent fuel to the total loaded fuel was changed from 5% to 20%. The required uranium enrichment for criticality becomes higher with the increasing of VF as well as the enlarging of the fraction of spent fuel in loaded fuel. The neutron spectra become harder with the augmenting of VF. The plutonium and minor actinides isotopes are produced more in the reactor. Copyright (C) 2015, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved.Article Citation - WoS: 11Citation - Scopus: 14Energy Multiplication and Fissile Fuel Breeding Limits of Accelerator-Driven Systems With Uranium and Thorium Targets(Pergamon-elsevier Science Ltd, 2015) Sahin, Sumer; Sarer, Basar; Celik, YurdunazThe study analyses the integral U-233 and Pu-239 breeding rates, neutron multiplication ratio through (n,xn)- and fission-reactions, heat release, energy multiplication and consequently the energy gain factor in infinite size thorium and uranium as breeder material in an accelerator driven systems (ADS), irradiated by a 1-GeV proton source. Energy gain factor has been calculated as M-energy = 1.67, 4.03 and 5.45 for thorium, depleted uranium (100% U-238) and natural uranium, respectively, where the infinite criticality values are k(infinity) = 0.40, 0.752 and 0.816. Fissile fuel material production is calculated as 53 Th-232(n,gamma)U-233, 80.24 and 90.65 U-238(n,gamma)Pu-239 atoms per incident proton, respectively. The neutron spectrum maximum is by similar to 1 MeV. Lower energy neutrons E < 1 MeV have major contribution on fissile fuel material breeding (>97.5%), whereas their share on energy multiplication is negligible (0.2%) for thorium, depleted uranium. Major fission events occur in the energy interval 1MeV < E < 50 MeV. Copyright (C) 2015, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved.Conference Object Citation - WoS: 2Citation - Scopus: 2Radiation Source Terms of Myrrha Reactor Components and Equipment(Pergamon-elsevier Science Ltd, 2016) Celik, Yurdunaz; Stankovskiy, Alexey; Engelen, Jeroen; Van den Eynde, Gert; Sarer, Basar; Sahin, SumerIn-vessel structural components of nuclear reactors are subject to prompt and residual neutron and photon activation. The MYRRHA fast spectrum facility, when operated in sub critical mode, suffers additional activation due to a wide range of energetic particles produced in the interactions of 600 MeV-primary protons with matter. The purpose of this work was to assess the source term (activation, heating and induced radiation level) of ex core equipment and components located inside the reactor vessel. Numerous stainless steel samples uniformly distributed inside the vessel have been used to simulate the activation of equipment in order to take into account the perturbation of the neutron spectrum caused by structural materials of components and equipment. The calculated quantities were prompt and residual activation, heating, radiation dose and radiation damage. The calculations were carried out with the ALEPH2 depletion code which invokes the MCNPX code for radiation transport. Copyright (C) 2016, Hydrogen Energy Publications, LLC. Published by Elsevier Ltd. All rights reserved.Editorial Preface To the Special Issue on "17th International Conference on Emerging Nuclear Energy Systems (icenes'2015), 4-8 October 2015, Istanbul, Turkey"(Pergamon-elsevier Science Ltd, 2016) Sahin, Sumer; Sahin, Haci Mehmet; Martinez-Val, Jose; Wu, Yican[No Abstract Available]
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