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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 Power flattening and minor actinide burning in a thorium fusion breeder(PERGAMON-ELSEVIER SCIENCE LTD, 2002) Sahin, S; Sahin, HM; Sozen, A; Bayrak, MA neutronic analysis has been performed for a thorium fusion breeder with a special task of burning minor actinides Np-237, Am-241, Am-243 and Cm-244 and production of U-233, Pu-238, Am-242m and Cm-245 for spacecraft application. Pu-238 is an important radioisotopic energy source for spacecraft generators. As potential nuclear fuels in the foreseeable future, U-233, Am-242m and Cm-245 would allow one to build extremely compact space reactors. Natural lithium has been selected as the coolant medium for the nuclear heat transfer out of the fuel zone. Minor actinides out of 5 and 10 units of LWRs per metre of blanket height have been mixed with ThO2. Higher fission rates in minor actinides enables one to realise a power flattening in the fissile zone over three years of plant operation by a gradual increase in the radial direction at start-up. This has significant advantages with respect to plant operation over the long term and also with respect to a uniform utilisation of the nuclear fuel in the fissile zone. After three years of plant operation, the net U-233 production is similar to300 kg per metre of blanket height. The Pu-238 yield is 21 and 41 kg for a waste actinide charge out of 5 and 10 units of LWRs per metre of blanket height, respectively, and the Cm-245 yield is 1.1 and 2 kg, respectively. The net Am-242m production is practically nil. With waste actinides out of 10 reactor units per metre of blanket height, the flattening of the nuclear heat production density in the fissile zone is almost perfect. Waste actinides out of five reactor units per metre of blanket height allow still an excellent power flattening. The quasi-constant power shape is saved over 36 months. (C) 2002 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.
