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Publication Metadata only Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium(2002) Yapici, Huseyin; Übeylï, Mustafa; Yalçin, Şenay; Yapici, Huseyin, Mühendislik Fakültesi, Erciyes Üniversitesi, Kayseri, Turkey; Übeylï, Mustafa, Teknik Eǧitim Fakültesi, Gazi Üniversitesi, Ankara, Turkey; Yalçin, Şenay, Fen-Edebiyat Fakültesi, Bahçeşehir Üniversitesi, Istanbul, TurkeyIn this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO2, ThC, UO2, UC, U3Si2 and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of 239Pu breeding capability among the uranium fuels whereas UO2 has the lowest one. 239Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and 233U production values are 0.125 and 0.140 in the blanket fueled with ThO2 and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm3 in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼ 1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼ 0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction. © 2002 Elsevier Science Ltd. All rights reserved. © 2008 Elsevier B.V., All rights reserved.Publication Metadata only Trellis decoding of some decomposable codes(Institute of Electrical and Electronics Engineers Inc., 2003) Altay, Gökmen; Yalçin, Şenay; Uçan, Osman Nuri; Kurnaz, S.; Ince, F.; Onbasioglu, S.; Altay, Gökmen, Department of Electrical and Electronic Engineering, Bahçeşehir Üniversitesi, Istanbul, Turkey; Yalçin, Şenay, Department of Mathematics, Bahçeşehir Üniversitesi, Istanbul, Turkey; Uçan, Osman Nuri, Department of Electrical and Electronic Engineering, Istanbul Üniversitesi, Istanbul, TurkeyThis paper presents a maximum likelihood identical minimal trellises of d4SP codes and it (ML) soft decision decoding scheme to implement Viterbi significantly reduces the decoding complexity by Algorithm for some constructed decomposable codes of performing iterations for each sub-trellis. Bit error rate Hamming distance four and show that the decoding performances of some d4SP codes are obtained in additive complexity is simple, hence, it may be employed in trellis- white Gaussian channel (AWGN) and also complexity of based decoders. A bit error rate performance of some decoding processes is analyzed in detail. Decomposable codes that were obtained employing the technique in AWGN channel are also presented. © 2015 Elsevier B.V., All rights reserved.Publication Metadata only Neutronics analysis of HYLIFE-II blanket for fissile fuel breeding in an inertial fusion energy reactor(2003) Şahín, Sümer; Yalçin, Şenay; Şahin, Haci Mehmet; Übeylï, Mustafa; Şahín, Sümer, Teknik Eǧitim Fakültesi, Gazi Üniversitesi, Ankara, Turkey; Yalçin, Şenay, Fen-Edebiyat Fakültesi, Bahçeşehir Üniversitesi, Istanbul, Turkey; Şahin, Haci Mehmet, Teknik Eǧitim Fakültesi, Gazi Üniversitesi, Ankara, Turkey; Übeylï, Mustafa, Teknik Eǧitim Fakültesi, Gazi Üniversitesi, Ankara, TurkeyA protective, 60 cm thick flowing liquid wall coolant is investigated as energy carrier, and fusile and fissile breeder medium in an inertial fusion energy (IFE) reactor. Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF4 and UF4) starting with 2 mol% up to 12 mol%. For a plant operation period of 30 years, radiation damage values were found as DPA=∼65 for 2 mol% heavy metal in the coolant, and remain practically constant with increasing heavy metal fraction, well below the presumable limit of DPA=100. Helium production values are calculated as ∼270 appm for 2 mol% heavy metal fraction, also being far below the limit value of 500 appm and remain at the same level with increasing heavy metal fraction. Such a flowing protective liquid wall extents the lifetime of the rigid first wall structure to a plant lifetime of 30 years. Fissionable metal salt in the flowing liquid enables one to breed high quality fissile fuel for external reactors by a self-sustaining tritium breeding for the fusion plant and increases plant power output. © 2002 Elsevier Science Ltd. All rights reserved. © 2005 Elsevier B.V., All rights reserved.Publication Metadata only Fissile fuel breeding with peaceful nuclear explosives(2003) Şahín, Sümer; Yalçin, Şenay; YIldIz, Kadir; Şahín, Sümer, Teknik Eǧitim Fakültesi, Gazi Üniversitesi, Ankara, Turkey; Yalçin, Şenay, Fen-Edebiyat Fakültesi, Bahçeşehir Üniversitesi, Istanbul, Turkey; YIldIz, Kadir, Aksaray Muhendislik Fakultesi, Niğde Ömer Halisdemir University, Nigde, TurkeyNeutron physics analysis of a dual purpose modified PACER concept has been conducted. A protective liquid droplet jet zone of 2 m thickness is considered as coolant, energy carrier, and fusile and fissile breeder. Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF 4 and UF4) starting by 2 up to 12 mol.%. The neutronic model assumed a 30 m radius underground spherical geometry cavity with a 1 cm thick SS-304 stainless steel liner attached to the excavated rock wall. By a self-sufficient tritium breeding of 1.05 with 5 mol.% ThF4, or 9 mol.% UF4 an excess nuclear fuel breeding rate of 1900 kg/year of 233U or 3000 kg/year 239Pu of extremely high isotopic purity can be realized. This precious fuel can be considered for special applications, such as spacecraft reactors or other compact reactors. The heavy metal constituents in jet zone acts as an energy amplifier, leading to an energy multiplication of M=1.27 or 1.65 for 5 mol.% ThF4, or 9 mol.% UF4, respectively. As an immediate result of the strong neutron attenuation in the jet zone, radiation damage with dpa<1.4 and He<7 ppm after a plant operation period of 30 years will be well below the damage limit values. The site could essentially be abandoned, or the cavity could be used as a shallow burial site for other qualified materials upon decommissioning. Finally, the totality of the site with all nuclear peripheral sections must be internationally safeguarded carefully. © 2003 Elsevier B.V. All rights reserved. © 2008 Elsevier B.V., All rights reserved.Publication Metadata only Characterization of PdAu thin films on oxidized silicon wafers: Interdiffusion and reaction(Elsevier, 2003) Yalçin, Şenay; Avci, Recep; Yalçin, Şenay, Fen-Edebiyat Fakültesi, Bahçeşehir Üniversitesi, Istanbul, Turkey; Avci, Recep, Department of Physics, Montana State University, Bozeman, United StatesPlasma-deposited thin films prepared at room temperature, ranging from 46 to 250 Å of PdAu on ∼45-50 Å Si-oxide and Si-oxynitride films grown on Si wafers were studied. Grazing incidence X-ray diffraction, X-ray reflectivity, and XPS depth profile techniques were used to characterize the thin films. A reactive interface involving Pd- and Au-silicides is formed, linking the thin film to the Si-oxide and Si-oxynitride films: a small fraction of Pd and Au atoms from PdAu migrate into the Si substrate, first penetrating the oxide layer, and the small fraction of Si atoms from the oxide layer migrate into the PdAu film and form a silicide interlayer consisting of a reactive interface made up of mixtures of Au- and Pd-silicides interspersed within the matrix of PdAu and substrate. The concentration profiles of these silicides have a maximum at the interface with decay on both sides. The density and the face-centered cubic (fcc) lattice parameter of the film are determined to be ∼13 ± 1 g/cm 3 and ∼4.004 ± 0.014 Å, respectively. The ideal film density is expected to be ∼15.5 g/cm 3 , which suggests substantial defect density and light material mixture, causing more than 13% reduction in the mass density of the film. © 2003 Elsevier Science B.V. All rights reserved. © 2019 Elsevier B.V., All rights reserved.Publication Metadata only Transient temperature and thermally induced stress distributions in a partly-circumferentially heated cylindrical workpiece(2004) Yapici, Huseyin; Yalçin, Şenay; Yapici, Huseyin, Mühendislik Fakültesi, Erciyes Üniversitesi, Kayseri, Turkey; Yalçin, Şenay, Fen-Edebiyat Fakültesi, Bahçeşehir Üniversitesi, Istanbul, TurkeyThis paper presents the numerical solutions of the transient temperature and thermally induced stress distributions in a partly-circumferentially heated cylindrical hollow workpiece (steel) with conjugate heat transfer. Outer surface of the workpiece is heated partly-circumferentially heat flux as its remainder outer surface is circumferentially cooled with fluid (water). Three phenomena have been considered as, (1) conduction inside the cylinder, (2) convection from the cylinder surface to the surrounding fluid, and (3) thermal stress produced by high temperature gradient inside the cylinder. The governing flow and energy equations have been solved numerically by using a control volume approach. The PHOENICS 3.2 and HEATING7 computer codes have been used for the numerical evaluation. The transient calculations have been performed individually for four fluid inlet velocities, ui = 0.005, 0.01, 0.015 and 0.020 m/s, until the system attains steady-state. The results of this study clearly demonstrate that the temperature contours in the low inlet velocity cases are more near to a symmetric case with respect to the y = 0 plane than that in the high inlet velocity cases, and the increment of the inlet velocity exponentially reduces the temperatures and thermally induced stresses in the workpiece. The effective thermal stress differences occurring in the workpiece can be significantly reduced by the high fluid inlet velocity. © 2008 Elsevier B.V., All rights reserved.Publication Metadata only Neutronic analysis of a high power density hybrid reactor using innovative coolants(Indian Academy of Sciences, 2005) Yalçin, Şenay; Übeylï, Mustafa; Acir, Adem; Yalçin, Şenay, Fen-Edebiyat Fakültesi, Bahçeşehir Üniversitesi, Istanbul, Turkey; Übeylï, Mustafa, Mühendislik Fakültesi, TOBB University of Economics and Technology, Ankara, Turkey; Acir, Adem, Teknik Eǧitim Fakültesi, Gazi Üniversitesi, Ankara, TurkeyIn this study, neutronic investigation of a deuterium-tritium (DT) driven hybrid reactor using ceramic uranium fuels, namely UC, UO2 or UN under a high neutron wall load (NWL) of 10 MW/m2 at the first wall is conducted over a period of 24 months for fissile fuel breeding for light water reactors (LWRs). New substances, namely, Flinabe or Li20Sn80 are used as coolants in the fuel zone to facilitate heat transfer out of the blanket. Natural lithium is also utilized for comparison to these two innovative coolants. Neutron transport calculations are performed on a simple experimental hybrid blanket with cylindrical geometry with the help of the SCALE 4·3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and an S8-P3 approximation. The investigated blanket using Flinabe or Li20Sn80 shows better fissile fuel breeding and fuel enrichment characteristics compared to that with natural lithium which shows that these two innovative coolants can be used in hybrid reactors for higher fissile fuel breeding performance. Furthermore, using a high NWL of 10 MW/m2 at the first wall of the investigated blanket can decrease the time for fuel rods to reach the level for charging in LWRs. © 2018 Elsevier B.V., All rights reserved.Publication Metadata only Numerical study on transient local entropy generation in pulsating turbulent flow through an externally heated pipe(Indian Academy of Sciences, 2005) Yapici, Huseyin; Genç, Gamze; Kayataş, Nesrin; Yalçin, Şenay; Yapici, Huseyin, Mühendislik Fakültesi, Erciyes Üniversitesi, Kayseri, Turkey; Genç, Gamze, Mühendislik Fakültesi, Erciyes Üniversitesi, Kayseri, Turkey; Kayataş, Nesrin, Mühendislik Fakültesi, Erciyes Üniversitesi, Kayseri, Turkey; Yalçin, Şenay, Mühendislik Fakültesi, Bahçeşehir Üniversitesi, Istanbul, TurkeyThis study presents an investigation of transient local entropy generation rate in pulsating turbulent flow through an externally heated pipe. The flow inlet to the pipe pulsates at a constant period and amplitude, only the velocity oscillates. The simulations are extended to include different pulsating flow cases (sinusoidal flow, step flow, and saw-down flow) and for varying periods. The flow and temperature fields are computed numerically with the help of the Fluent computational fluid dynamics (CFD) code, and a computer program developed by us by using the results of the calculations performed for the flow and temperature fields. In all investigated cases, the irreversibility due to the heat transfer dominates. With the increase of flow period, the highest levels of the total entropy generation rates increase logarithmically in the case of sinusoidal and saw-down flow cases whereas they are almost constant and the highest total local entropy is also generated in the step case flow. The Merit number oscillates periodically in the pulsating flow cases along the flow time. The results of this study indicate that flow pulsation has an adverse effect on the ratio of the useful energy transfer rate to the irreversibility rate. © 2018 Elsevier B.V., All rights reserved.Publication Metadata only Belief propagation decoding of some decomposable linear block codes(2005) Altay, Gökmen; Yalçin, Şenay; Altay, Gökmen, Department of Electrical and Electronic Engineering, Bahçeşehir Üniversitesi, Istanbul, Turkey; Yalçin, Şenay, Faculty of Engineering, Bahçeşehir Üniversitesi, Istanbul, TurkeyWe implement the Belief Propagation Algorithm onto some decomposable linear block codes and obtain bit error rate performances for some of the decomposable codes over Additive White Gaussian Noise channel. A comparison between the Belief Propagation Algorithm and the Viterbi Algorithm is also performed with respect to the obtained error performances. ©2005 IEEE. © 2008 Elsevier B.V., All rights reserved.Publication Metadata only Utilization of refractory metals and alloys in fusion reactor structures(Springer Science and Business Media Deutschland GmbH, 2006) Übeylï, Mustafa; Yalçin, Şenay; Übeylï, Mustafa, Mechanical Engineering, TOBB University of Economics and Technology, Ankara, Turkey; Yalçin, Şenay, Computer Engineering, Bahçeşehir Üniversitesi, Istanbul, TurkeyIn design of fusion reactors structural material selection is very crucial to improve reactor's performance. Different types of materials have been proposed for use in fusion reactor structures. Among these materials refractory metals and alloys having capability to withstand high temperatures and high neutron wall loads have been considered to get high power density in fusion reactors. However these materials have insufficient technological database and are very expensive compared to steels. In addition to that except chromium and some chromium alloys they show no low activation property. This study gives an overview of potential of refractory metals and alloys for possible use in fusion reactors. © Springer Science+Business Media Inc. 2006. © 2019 Elsevier B.V., All rights reserved.
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