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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]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: 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.Article Citation - WoS: 22Citation - Scopus: 26Hydrogen Hazard and Mitigation Analysis in Pwr Containment(Pergamon-elsevier Science Ltd, 2013) Sahin, Sumer; Sarwar, Mohammad SohailThis paper describes the analytical results for the estimation of hydrogen concentration in the containment atmosphere based on zirconium oxidation reaction following a severe accident. The analysis provides useful information about the potential challenge of local hydrogen accumulation in the containment, which may be used to reduce the hydrogen detonation risk and to design the capacity and arrangements of mitigation measures. The containment analysis has been performed using computer code COGAP which uses the scenario of loss of coolant accident. The behavior of pressure and hydrogen concentrations in containment as a function of time under the severe accident condition is presented in graphical form. The mitigation measures (recombiners) are essential to maintain containment atmosphere in the safe stable conditions. A hydrogen control system is to mitigate the hydrogen risk by comparing results from a reference accident sequence with and without recombiners. This comparison show that combustible gas occur in few local areas in the containment for a limited time span and hydrogen concentration is reduced significantly with the use of recombiners. (C) 2013 Elsevier Ltd. All rights reserved.Article Citation - WoS: 3Citation - Scopus: 4Experimental Evaluation of Surveillance Capsule Assemblies for Life Assessment of Chasnupp Unit-1 Reactor Pressure Vessel(Pergamon-elsevier Science Ltd, 2016) Sahin, Sumer; Saeed, AsimNeutron flux and energy spectrum were determined at the surface of three in-vessel Surveillance Capsule Assemblies (SCAs) removed from CHASNUPP Unit-1 after 2nd, 4th, and 9th fuel cycles for the life assessment of reactor pressure vessel belt line region. Dosimetry data were measured from radiometric sensors irradiated in base material section of SCAs. Fast neutron flux (E > 1.0 MeV) was best estimated at the surface of three SCAs corresponding to the center of C-1 core using the least square method by employing LSL-M2 package. These results were compared with fast neutron flux calculated using DOT3.5 code and both results are within good agreement of +/- 20% acceptance criteria as described in Regulatory Guide 1.190. Therefore, calculational model was validated by dosimetry evaluation and these results can be used in the life assessment of CHASNUPP Unit-1 pressure vessel belt line region. (C) 2015 Elsevier Ltd. All rights reserved.Article Citation - WoS: 21Citation - Scopus: 26LIFE hybrid reactor as reactor grade plutonium burner(Pergamon-elsevier Science Ltd, 2012) Sahin, Sumer; Sahin, Haci Mehmet; Acir, AdemThe early version of the conceptual modified design of the Laser Inertial Confinement Fusion Fission Energy (LIFE) engine consists of a spherical fusion chamber of 5 m diameter, surrounded by a multi-layered blanket. The first wall is made of 2 cm thick ODS and followed by a Li17Pb83 zone (2 cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2 cm) from the FLIBE molten salt zone (50 cm), containing fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a constant fusion driver power of 500 MWth, in S-8-P-3 approximation using 238-neutron groups. Reactor grade (RG) plutonium carbide fuel in form of TRISO particles with volume fractions of 2%, 3%, 4%, 5% and 6% have been dispersed homogenously in the FLIBE coolant. Tritium breeding ratio (TBR) values per incident fusion neutron for the above cited cases start with TBR = 1.35, 1.52, 1.73, 2.02 and 2.47, respectively. With the depletion of fissionable RG-Pu isotopes, TBR decreases gradually. At startup, higher fissionable fuel content in the molten salt leads to higher blanket energy multiplication, namely M-0 = 3.8, 5.5, 7.7, 10.8 and 15.4 with 2%, 3%, 4%, 5% and 6% TRISO volume fraction, respectively. Calculations have led to very high burn up values (>400,000 MD.D/MT). TRISO particles can withstand such high burn ups. Such high burn ups would lead to drastic reduction of final nuclear waste per unit energy production. (C) 2012 Elsevier Ltd. All rights reserved.Article Citation - WoS: 6Citation - Scopus: 15Investigation of a Gas Turbine-Modular Helium Reactor Using Reactor Grade Plutonium With 232th and 238u(Pergamon-elsevier Science Ltd, 2016) Sahin, Sumer; Erol, Ozgur; Sahin, Haci MehmetUtilization of natural uranium (nat-U) and thorium as fertile fuels has been investigated by in a Gas Turbine - Modular Helium Reactor (GTMHR) using reactor grade plutonium as driver fuel. A neutronic analysis for the full core reactor was performed by using MCNP5 with ENDF/B-VI cross-section library. Different mixture ratios were tested in order to find the appropriate mixture ratio of fertile and fissile fuel particles that gives a comparable k(eff) value of the reference uranium fuel. Time dependent calculations were performed by using MONTEBURN2.0 with ORIGEN2.2 for each selected mixture. Different parameters (operation time, burnup value, fissile isotope change, etc.) were subject of performance comparison. The operation time and burnup values were close to each other with nat-U and thorium, namely 3205 days and 176 GWd/MTU for the former and 3175 days 181 GWd/MTU for the latter fertile fuel. In addition, the fissile isotope amount changed from initially 6940.1 kg-4579.2 kg at the end of its operation time for nat-U. These values were obtained for thorium as 6603.3 kg-4250.2 kg, respectively. (C) 2016 Elsevier Ltd. All rights reserved.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 Editor's Report on Nurer2012, the Iii. International Conference on Nuclear and Renewable Energy Resources, Istanbul, Turkiye (20-23rd May 2012)(Pergamon-elsevier Science Ltd, 2013) Sahin, Sumer[No Abstract Available]
