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Huseyin Yapici - One of the best experts on this subject based on the ideXlab platform.
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Time-Dependent Neutronic Analysis of a Power-Flattened Gas Cooled Accelerator Driven System Fuelled with Thorium, Uranium, Plutonium, and Curium Dioxides TRISO Particles
Science and Technology of Nuclear Installations, 2016Co-Authors: Gizem Bakir, Gamze Genc, Huseyin YapiciAbstract:This study presents the power flattening and time-dependent Neutronic Analysis of a conceptual helium gas cooled Accelerator Driven System (ADS) loaded with TRISO (tristructural-isotropic) fuel particles. Target material is lead-bismuth eutectic (LBE). ThO2, UO2, PuO2, and CmO2TRISO particles are used as fuel. PuO2and CmO2fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the Neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power (Pth) of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g233U and 103.6 g239Pu per day can be produced. The considered system has a high Neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization.
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Time-dependent Neutronic Analysis for High Level Waste Transmutation in a Fusion-driven Transmuter
Journal of Fusion Energy, 2007Co-Authors: Huseyin Yapici, Nesrin Demir, Gamze GencAbstract:This study presents time-dependent transmutations of high-level waste (HLW) including minor actinides (MAs) and long-lived fission products (LLFPs) in the fusion-driven transmuter (FDT) that is optimized in terms of the Neutronic performance per fusion neutron in our previous study. Its blanket has two different transmutation zones (MA transmutation zone, TZMA, and LLFP transmutation zone, TZFP), located separately from each other. High burn-up pressured water reactor (PWR)-mixed oxide (MOX) spent fuel is used as HLW. The time-dependent transmutation analyses have been performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m2. The effective half-lives of the MA and LLFP nuclides can be shortened significantly in the considered FDT while substantial electricity is produced in situ along the OP.
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On the Neutronic Performance of Hylife-II Reactor Fuelled with Carbide Fuels
Journal of Fusion Energy, 2006Co-Authors: Huseyin Yapici, Mustafa ÜbeyliAbstract:In this study, Neutronic Analysis of the HYLIFE-II reactor was investigated by inserting fuel rods containing UC or mixed ThC–UC into reflector zone partially. Four different coolants, namely, flibe, helium, natural lithium, and light water were considered in the fissile fuel breeding zone for comparison. Neutron transport calculations per incident (D,T) fusion neutron were performed by using the code Scale 4.3 under resonance-effect and resonance-free conditions. Numerical results pointed out that replacing the reflector zone by fissile fuel breeding zone even with a thickness of 14 cm improved the Neutronic performance remarkably with respect to energy amplification and fissile fuel breeding.
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Neutronic Analysis of denaturing plutonium in a thorium fusion breeder and power flattening
Energy Conversion and Management, 2005Co-Authors: Huseyin Yapici, Mustafa BayrakAbstract:Abstract The purpose of this study is to denature nuclear weapon grade quality plutonium in a thorium fusion breeder. Ten fuel rods containing the mixture of ThO 2 and PuO 2 are placed in a radial direction in the fissile zone where ThO 2 is mixed with variable amounts of PuO 2 to obtain a quasi-constant nuclear heat production density. The plutonium composition volume fractions in the fuel rods are gradually increased from 0.1% to 1% by 0.1% increments. The fissile fuel zone is cooled with four various coolants with a volume fraction ratio of 1 ( V coolant / V fuel = 1). These coolants are helium gas, flibe “Li 2 BeF 4 ”, natural lithium and eutectic lithium “Li 17 Pb 83 ”. Nuclear weapon grade quality 239 Pu in the fuel composition is denatured due to the accumulation of the 240 Pu isotope in the fissile zone after 18 months of plant operations. Under a first wall fusion neutron current load of 2.222 × 10 14 (14.1 MeV n/cm 2 s), which corresponds to 5 MW/m 2 , by a plant factor of 100%, at the end of the plant operation, the fissile fuel enrichment quality between 6.0% and 10% is obtained depending on the coolant types. During the plant operation, the tritium breeding ratio (TBR) should be at least 1.05. In the selected blanket, only the flibe coolant is already self sustaining at start up. The TBR increases steadily due to the higher neutron multiplication rate during the plant operation period. The highest TBR is obtained for the eutectic lithium coolant 1.4035, followed by the flibe coolant 1.3095, helium gas coolant 1.2172 and natural lithium coolant 1.0553 at the end of the operation period of 48 months. The energy multiplication factor M changed between 2.1731 and 6.6241 depending on coolant type during the operation period. The peak to average fission power density ratio Γ in the blanket decreases by ∼15%, which allows a more uniform power generation in the fissile zone. The isotopic percentage of 240 Pu reaches higher than 5% in all coolant types. This is very important for international safety.
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Neutronic Analysis of prometheus reactor fueled with various compounds of thorium and uranium
Annals of Nuclear Energy, 2002Co-Authors: Huseyin Yapici, Mustafa Übeyli, Senay YalçinAbstract:In 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.
Muhammad Ilham - One of the best experts on this subject based on the ideXlab platform.
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comparison of uranium plutonium nitride u pu n and thorium nitride th n fuel for 500 mwth gas cooled fast reactor gfr long life without refueling
International Journal of Energy Research, 2018Co-Authors: Ratna Dewi Syarifah, Zaki Suud, Khairul Basar, Dwi Irwanto, Sandro Clief Pattipawaej, Muhammad IlhamAbstract:Summary Comparison of uranium plutonium nitride and thorium nitride fuel for 500 MWth gas-cooled fast reactor has been done. Gas-cooled fast reactor is one type of generation IV reactor that can be operated in high temperature. Due to the high temperature, it can be used in hydrogen production. In this study, we compare the Neutronic Analysis of two fuel types, ie, uranium nitride fuel (U,Pu)N and thorium nitride fuel (Th,U233)N. The Neutronic calculation uses SRAC2006 code system, and the data libraries use JENDL4.0. First, the fuel pin calculation (PIJ calculation) has been done to take the macro data that are used in CITATION calculation. Both uranium and thorium fuel use heterogeneous configuration with 3 variation fuel in the core. F1 is located in the central core, F2 middle core, and F3 outer core. The variation of fuel fraction is 40% until 65%, cladding 10%, and coolant 25% until 40%. For (U,Pu)N fuel, the diameter of the active core is 220 cm, and the height of the active core is 110 cm. And for (Th,U233)N, the diameter of the active core is 250 cm and the height of the active core is 150 cm. The reflector radial-axial width is 50 cm. For uranium plutonium nitride fuel, the type of the fuel in one core is varied; ie, F1 is 8%, F2 is 10%, and F3 is 12%. For thorium nitride fuel, the type of the fuel in one core also varied; ie, F1 is 7.8%, F2 is 8%, and F3 is 8.8%. The optimum value of thorium nitride is when fuel fraction of region F1 = 60%, F2 = 57.5%, F3 = 60%, the burn up time up to 20 years without refueling, max k-eff value is 1.0109, and max excess reactivity value is 1.08%. Neutronic Analysis shows that both uranium and thorium fuel have excess reactivity value less than 2% but thorium fuel has excess reactivity less than uranium nitride fuel. Uranium fuel has better breeding capability than thorium fuel. Therefore, it is better to use uranium fuel for fast reactor like GFR, which has high breeding capability.
Mustafa Bayrak - One of the best experts on this subject based on the ideXlab platform.
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Neutronic Analysis of denaturing plutonium in a thorium fusion breeder and power flattening
Energy Conversion and Management, 2005Co-Authors: Huseyin Yapici, Mustafa BayrakAbstract:Abstract The purpose of this study is to denature nuclear weapon grade quality plutonium in a thorium fusion breeder. Ten fuel rods containing the mixture of ThO 2 and PuO 2 are placed in a radial direction in the fissile zone where ThO 2 is mixed with variable amounts of PuO 2 to obtain a quasi-constant nuclear heat production density. The plutonium composition volume fractions in the fuel rods are gradually increased from 0.1% to 1% by 0.1% increments. The fissile fuel zone is cooled with four various coolants with a volume fraction ratio of 1 ( V coolant / V fuel = 1). These coolants are helium gas, flibe “Li 2 BeF 4 ”, natural lithium and eutectic lithium “Li 17 Pb 83 ”. Nuclear weapon grade quality 239 Pu in the fuel composition is denatured due to the accumulation of the 240 Pu isotope in the fissile zone after 18 months of plant operations. Under a first wall fusion neutron current load of 2.222 × 10 14 (14.1 MeV n/cm 2 s), which corresponds to 5 MW/m 2 , by a plant factor of 100%, at the end of the plant operation, the fissile fuel enrichment quality between 6.0% and 10% is obtained depending on the coolant types. During the plant operation, the tritium breeding ratio (TBR) should be at least 1.05. In the selected blanket, only the flibe coolant is already self sustaining at start up. The TBR increases steadily due to the higher neutron multiplication rate during the plant operation period. The highest TBR is obtained for the eutectic lithium coolant 1.4035, followed by the flibe coolant 1.3095, helium gas coolant 1.2172 and natural lithium coolant 1.0553 at the end of the operation period of 48 months. The energy multiplication factor M changed between 2.1731 and 6.6241 depending on coolant type during the operation period. The peak to average fission power density ratio Γ in the blanket decreases by ∼15%, which allows a more uniform power generation in the fissile zone. The isotopic percentage of 240 Pu reaches higher than 5% in all coolant types. This is very important for international safety.
Shokufe Forughi - One of the best experts on this subject based on the ideXlab platform.
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Neutronic Analysis for tehran research reactor mixed core
Progress in Nuclear Energy, 2012Co-Authors: A Lashkari, Hossein Khalafi, Mohammad S Mirvakili, Shokufe ForughiAbstract:In this research, Neutronic calculation of current low enriched uranium control fuel elements replacement with high enriched uranium control fuel elements in the reference core of Tehran Research Reactor (TRR) has been investigated and the results of calculations are compared with the TRR Neutronic safety criteria. Results show that all Neutronic parameters of the reference and each mixed-core are lower than the safety criteria. Nuclear reactor Analysis codes including MTR_PC package and MCNP5 were employed to carry out these calculations.
Mohammad S Mirvakili - One of the best experts on this subject based on the ideXlab platform.
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Neutronic Analysis for tehran research reactor mixed core
Progress in Nuclear Energy, 2012Co-Authors: A Lashkari, Hossein Khalafi, Mohammad S Mirvakili, Shokufe ForughiAbstract:In this research, Neutronic calculation of current low enriched uranium control fuel elements replacement with high enriched uranium control fuel elements in the reference core of Tehran Research Reactor (TRR) has been investigated and the results of calculations are compared with the TRR Neutronic safety criteria. Results show that all Neutronic parameters of the reference and each mixed-core are lower than the safety criteria. Nuclear reactor Analysis codes including MTR_PC package and MCNP5 were employed to carry out these calculations.