Yazar "Yapici, H" seçeneğine göre listele
Listeleniyor 1 - 8 / 8
Sayfa Başına Sonuç
Sıralama seçenekleri
Öğe Analysis of the rejuvenation performance of hybrid blankets by using uranium fuels (UN, UC, UO2, U3Si2) and different coolants for various volume fraction(PERGAMON-ELSEVIER SCIENCE LTD, 2000) Yapici, H; Ipek, O; Ozceyhan, V; Erisen, AThe possibility of nuclear fuel rejuvenation in fusion reactors is investigated for different fuels and coolants. Neutronic performances of the deuterium-tritium (D-T) driven hybrid blankets, fuelled with UN, UC, UO2 and U3Si2, in four different cases, are investigated under first wall load of the 5 MW/m(2.) The fissile fuel zone considered to be cooled with four coolants: air, flibe (Li2BeF4), natural lithium and eutectic lithium (Li17Pb83) With volume fraction ratio of 29.5, 45.5 and 62.56%. The behaviour of the fuels mentioned above are observed during 48 months for discrete time intervals of Delta t = 15 days and by a plant factor (PF) of 75%. At the end of the operation time, calculations have shown that cumulative fissile fuel enrichment (CFFE) values have varied between 3.80 and 8.1% depending on the fuel, volume fraction and coolant type. The best enrichment performance is obtained in flibe (Li2BeF4) coolant blankets, followed by Eutectic lithium (Li17Pb83), air whereas natural lithium coolant shows a poor rejuvenation performance in all fuels. CFFE reach maximum value (8.1%) in UO2 fuelled blanket (in Row #1) and Li2BeF4 coolant that volume fraction is 62.5% after 48 months. The lowest CFFE value (3.80%) is in U3Si2 fuelled blanket (in Rows #6 and 7) and natural lithium coolant that volume fraction is 62.56% at the end of the operation period.The enrichment would be sufficient for LWR reactor. The best tritium breeding ratio (TBR) is obtained in U3Si2 fuelled blanket with natural lithium coolant, and followed by UC, UO2. UN with the same coolant. At the beginning of the operation, TBR values were 1.459, 1.502 and 1.554 in U3Si2 fuelled blanket with natural Lithium coolant 1.414, 1.474 and 1.547 in UC fuelled blanket with natural lithium coolant for volume fraction of 29.5, 45.5 and 62.56%, respectively. At the end of the operation, TBR reach 1.511. 1.559 and 1.613 in U3Si2 fuelled blanket and 1.467, 1.532 and 1.609 in UC fuelled blanket for volume fraction of 29.5, 45.5 and 62.56%, respectively. TBR values are higher than unity. Therefore, investigated hybrid blanket is self-sufficient for all fuel mixture and coolants. The isotopic percentage of Pu-240 is higher than 5% in all modes with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is a very important safety factor. The isotopic percentage of Pu-240 is lower than 5% in all blanket with air, natural lithium, and eutectic lithium coolant. In these modes, operation period must be increased for safety. (C) 2000 Elsevier Science Ltd. All rights reserved.Öğe Investigation of neutronic potential of a moderated (D-T) fusion driven hybrid reactor fueled with thorium to breed fissile fuel for LWRs(PERGAMON-ELSEVIER SCIENCE LTD, 2000) Yapici, H; Sahin, N; Bayrak, MThe potential of a moderated hybrid reactor fueled with ThC2 or ThF4 is investigated by a PF (plant factor) 75% under a first wall fusion neutron current load of 5 MW/m(2) LWR (Light Water Reactor) fuel rods containing ThC2 or ThF4 are replaced in the fissile fuel zone of the hybrid reactor. It is considered that gas (He or CO2), or flibe (Li2BeF4), or natural lithium is the coolant. The behaviour of the neutronic potential is observed for four years. At the end of the operation period, the Cumulative Fissile Fuel Enrichment (CFFE) values varied between 3.55 and 7% depending on the fuel and coolant type. Calculations show that the best neutronic performance is obtained with Flibe, followed by air and natural lithium coolants. After 48 months, the maximum CFFE value is 7% in the ThF4 fuel and flibe coolant mode, and the lowest CFFE value 3.55% is in the ThC2 fuel and natural lithium coolant mode. Consequently, these enrichments would be sufficient for LWRs. The Tritium Breeding Ratio (TBR) values are greater than 1.05 for all investigated natural lithium coolant modes, and the investigated hybrid reactor is self-sufficient in the tritium required for the D,T fusion driver in these modes during the operation period. The blanket energy multiplication factor M, varies between 2.45 and 3.68 depending on the type of fuel and coolant at the end of the operation period. At the same time, the peak-to-average fission power density ratio decreases by similar to 25%. The lowest radial neutron leakage out of the blanket is in the blanket with the flibe coolant modes. (C) 1999 Elsevier Science Ltd. All rights reserved.Öğe Neutronic analysis of denaturing plutonium in a thorium fusion breeder and power flattening(PERGAMON-ELSEVIER SCIENCE LTD, 2005) Yapici, H; Bayrak, MThe purpose of this study is to denature nuclear weapon grade quality plutonium in a thorium fusion breeder. Ten fuel rods containing the mixture of ThO2 and PuO2 are placed in a radial direction in the fissile zone where ThO2 is mixed with variable amounts of PuO2 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 "Li2BeF4". natural lithium and eutectic lithium "Li-17Pb83". 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 IS months of plant operations. Under a first wall fusion neutron. current load of 2.222 x 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 1.0551 at the end of the operation period of 48 months. The energy multiplication factor M changed between 2.17331 and 6.6241 depending on coolant type during the operation period. The peak to average fission power density ratio F in the blanket decreases by similar to 15 %, which allows a more uniform power generation in the fissile zone. The isotopic percentage of Pu-240 reaches higher than 5 % in all coolant types. This is very important for international safety. (C) 2004 Elsevier Ltd. All rights reserved.Öğe Neutronic performance of proliferation hardened thorium fusion breeders(ELSEVIER SCIENCE SA, 2001) Sahin, S; Yapici, H; Sahin, NProduction of denaturated fissile fuel in a thorium fusion breeder has been investigated by mixing the fertile fuel with natural-UO2 and LWR (light water reactors) spent nuclear fuel. Four different coolants (pressurised helium, Flibe 'Li2BeF4', natural lithium and Li17Pb83 eutectic) are selected for the nuclear heat transfer. In order to obtain a power flattening in the fissile-fertile zone, the UO2- or the spent fuel-fraction in the mixed oxide (MOX) fuel has been gradually increased in radial direction. Power plant operation periods between 13 and 29 months are evaluated to achieve a fissile fuel enrichment similar to4%, under a first-wall fusion neutron energy current of 5 MW m(-2) (plant factor of 100%). For a plant operation over 4 years, enrichment grades between 6.0 and 11.5% are calculated for the investigated MOX fuel and coolant compositions. The U-238 component of natural-UO2 can provide a limited proliferation hardening only for the U-233 component, whereas, the homogenous mixture of ThO2 with a small quantity of (>4%) LWR spent nuclear fuel remains all-over non-prolific for both, uranium and plutonium components. (C) 2001 Elsevier Science B.V. All rights reserved.Öğe Numerical neutronic analysis of a natural lithium cooled fusion breeder fueled with UO2(KING FAHD UNIV PETROLEUM MINERALS, 2000) Yapici, H; Ozceyhan, VThe fissile breeding capability of a catalyzed-(D,D) and (D,T) fusion-fission (hybrid) reactor, fueled with spent UO2, is analyzed under first wall fusion neutron load of 5 MW/m(2) to provide nuclear fuel for LWRs. This can be a prospective alternative to the existing methods of nuclear fuel enrichment. Lithium (Li) and lithium beryllium mixtures are chosen for the nuclear heat transfer out from the fissile fuel-breeding zone. The behavior of the UO2 fuel is observed during the 48 months for discrete time intervals of Delta t = 15 days and over a plant factor of 75%. Calculations show that a residence time of 12 to 42 months in possible for spent UO2 kept inside a fusion-fission reactor so as to accumulate cumulative fissile fuel enrichment values that would meet an acceptable quality level for deployment of the irradiated UO2 as fuel in LWRs. Enrichment grades between 4.5% and 6.5% can be achieved during a plant operation over four years depending on the type of fusion driver and coolant. In all types, the tritium breeding ratio (TBR) exceeds unity. Therefore, the blanket is self-sufficient with respect to tritium breeding. Mathematical models have been established for important nuclear engineering criteria depending on the type of fusion driver and coolant in spent fuel rejuvenation.Öğe Potential of a fusion-fission hybrid reactor using uranium for various coolants to breed fissile fuel for LWRs(PERGAMON-ELSEVIER SCIENCE LTD, 1999) Yapici, H; Curuttu, I; Ozceyhan, V; Kirbiyik, MIn this study, neutronic performances of the (D,T) driven hybrid blankets, fuelled with UC2 and UF4, are investigated under first wall load of 5 MW/m(2). The fissile fuel zone is considered to be cooled with three coolants: gas (He or CO2), flibe (Li2BeF4), and natural lithium. The behaviour of the UC2 and UF4 fuels are observed during 48 months for discrete time intervals of Delta t = 15 days and by a plant factor of 75%. At the end of the operation time, calculations have shown that Cumulative Fissile Fuel Enrichment (CFFE) values varied between 5 and 8.5% depending on the fuel and coolant type. The best enrichment performance is obtained in UF4 fuelled blanket with flibe coolant, followed by gas and natural lithium coolant. CFFE reaches maximum value (8.51%) in UF4 fuelled blanket (in row #1) and flibe coolant mode after 48 months. The lowest CFFE value (4.71%) is in UC2 fuelled blanket (in row #8) and natural lithium coolant at the end of the operation period. This enrichment would be sufficient for LWR reactor. At the beginning of the operation, tritium breeding ratio (TBR) values were 1.090, 1.3301 and 1.2489 in UC2 fuelled blanket and 1.0772, 1.2433 and 1.1533 in UF4 fuelled blanket for flibe, natural lithium and gas coolant, respectively. At the end of the operation, TBR reach 1.1820, 1.3983 and 1.3138 in UC2 fuelled blanket and 1.2041,1.3266 and 1.2407 in UF4 fuelled blanket for flibe, natural lithium and gas coolant, respectively. Nuclear quality of the plutonium increases linearly during the operation period. The isotopic percentage of Pu-240 is higher than 5% in UF4 and UC2 fuel with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is very important factor for safeguarding. The isotopic percentage of Pu-240 is lower than 5% in UC2 fuel with gas and natural lithium coolant. In these modes, operation period must be increased to safeguarding. (C) 1999 Elsevier Science Ltd. All rights reserved.Öğe Proliferation hardening and power flattening of a thorium fusion breeder with triple mixed oxide fuel(PERGAMON-ELSEVIER SCIENCE LTD, 2001) Sahin, S; Ozceyhan, V; Yapici, HThe proliferation hardening of the U-233 fuel in a thorium fusion breeder has been realised successfully with a homogenous mixture of ThO2, natural-UO2 and CANDU spent nuclear fuel in the form of a triple mixed oxide (TMOX) fuel. The new U-233 component will be successfully hardened against proliferation with the help of the U-238 component in the natural-UO2 and spent fuel. The plutonium component remains non-prolific through the presence of the Pu-240 isotope in the spent CANDU fuel due to its high spontaneous fission rate. A (D,T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. A quasi-constant power density in the fissile zone has been achieved by reducing the ThO2 component in the rods continuously in the radial direction (from 91 down to 64%). Three different coolants (pressurised helium, natural lithium and Li17Pb83 eutectic) are selected for the nuclear heat transfer out of the fissile fuel breeding zone with a volume ratio of V-coolant/V-fuel = 1 in the fissile zone. The fissile fuel breeding occurs through the neutron capture reaction in the Th-232 (ThO2), in the U-238 (natural-UO2 and CANDU spent fuel) isotopes. The fusion breeder increases the nuclear quality of the spent fuel, which can be defined with the help of the cumulative fissile fuel enrichment (CFFE) grade of the nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials (U-233+U-235+Pu-239+Pu-241) in the TMOX fuel. Under a first-wall fusion neutron current load of 10(14) (14.1 MeVn/cm(2) s), corresponding to 2.25 MW/m(2) and by a plant factor of 100%, the TMOX fuel can achieve an enrichment degree of similar to1% after similar to 12-15 months. A longer irradiation period (similar to 30 months) increases the fissile fuel enrichment levels of the TMOX towards much higher degrees (similar to2%), opening new possibilities for utilisation in advanced CANDU thorium breeders. The selected TMOX fuel remains non-prolific over the entire period for both uranium and plutonium components. This is an important factor with regard to international safeguarding. (C) 2000 Elsevier Science Ltd. All rights reserved.Öğe Spent mixed oxide fuel rejuvenation in fusion breeders(ELSEVIER SCIENCE SA, 1999) Sahin, S; Yapici, H; Bayrak, MA fusion breeder is presented for the rejuvenation of spent nuclear fuel. A (D, T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing ten rows in radial direction, covers the cylindrical fusion plasma chamber. The first three fuel rod rows contain Canadian deuterium uranium (CANDU) reactor spent nuclear fuel which was used down to a total enrichment grade of 0.418%. The following seven fuel rod rows contain light water reactor (LWR) spent nuclear fuel, which was used down to a total enrichment grade of 2.17%. This allows a certain degree of fission power flattening. Fissile zone is cooled with pressurised helium gas with volume ration of V-coolant/V-fuel = 2 in the fissile zone. Spent fuel rejuvenation occurs through the neutron capture reaction in U-238. The new fissile material increases the nuclear quality of the spent fuel which can be described as the cumulative fissile fuel enrichment (CFFE) grade of the nuclear fuel which is the sum of the isotopic ratios of all fissile material (U-235 + (PU)-P-239 + (PU)-P-241) in the mixed oxide (MOX) fuel. Under a first-wall fusion neutron current load of 10(14) (14.1-MeV n/cm(2) s), corresponding to 2.25 MW/m(2) and by a plant factor of 100%, the CANDU spent fuel can achieve an enrichment degree of 1% after similar to 7 months, suitable for reutilization in a CANDU reactor. LWR spent fuel requires > 15 months to reach an enrichment grade similar to 3.5%, suitable for reutilization in a LWR. A longer rejuvenation period (up to 48 months) increases the fissile fuel enrichment levers of the spent fuel reactor to much higher degrees (> 3% for CANDU spent fuel and over 5% for LWR spent fuel), opening possibilities an increased burn-up in critical reactors and a re-utilization in multiple cycles. (C) 1999 Elsevier Science S.A. All rights reserved.
