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    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, A
    The 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.
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    Numerical neutronic analysis of a natural lithium cooled fusion breeder fueled with UO2
    (KING FAHD UNIV PETROLEUM MINERALS, 2000) Yapici, H; Ozceyhan, V
    The 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.
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    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, M
    In 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.
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    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, H
    The 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.