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Öğe An assessment of thorium and spent LWR-fuel utilization potential in CANDU reactors(PERGAMON-ELSEVIER SCIENCE LTD, 2004) Sahin, S; Sahin, HM; Alkan, M; Yildiz, KA neutronic analysis has been performed to assess a prospective utilization of light water reactor (LWR) spent fuel in Canada deuterium uranium (CANDU) reactors mixed with thoria (ThO2). The study is conducted for mixture grades with 50%, 60% and 100% LWR spent fuel and 50%, 40% and 0% thoria, respectively. Burn-up grades are evaluated for alternative fuels to reach a bundle criticality of k(infinity) = 1.06, which are calculated as similar to28,000, similar to14;000, similar to8000 and 8800 MW d/MT with 100%, 60% and 50% LWR spent fuel content and for natural uranium fuelled CANDU after plant operation periods of 690, 340, 200 and 205 days, respectively. The presence of even plutonium isotopes with higher neutron absorption cross sections in the LWR spent fuel obliges starting with a higher cumulative fissile inventory in the initial charge compared to natural uranium fuel. Extended utilization of worldwide disposed spent nuclear LWR fuel in CANDU reactors in a symbiotic system opens prospects with respect to environmental concerns as well as to energy economics. After separation of the fission products, further utilization of the actinides in nuclear waste becomes possible as a valuable nuclear fuel. (C) 2003 Elsevier Ltd. All rights reserved.Öğe Effects of spectral shifting in an inertial confinement fusion system(CARL HANSER VERLAG, 2005) Sahin, S; Sahin, HM; Yildiz, K; Acir, AThe main objective is to study the effects of spectral shifting in an inertial confinement system for kT/shot energy regime on the breeding performance for tritium and for high quality fissile fuel. A protective liquid droplet jet zone of 2 m thickness is used as coolant, energy carrier and breeder Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF4 or UF4) starting by 2 moles% up to 12 moles%. Spectrum softening within the inertial confinement system reduces the tritium production ratio (TBR) in the protective coolant to a lower level than unity. However additional tritium production in the (Li2DT)-Li-6 zone of the system increases TBR to values above unity and allows a continuous operation of the power plant with a self-sustained fusion fuel supply. By modest fusion fuel burn efficiencies (40 to 60%) and with a few mol. % of heavy metal salt in the coolant in form of ThF4 or % UF4, a satisfactory TBR of > 1.05 can be realized. In addition to that, excess fissile fuel of extremely high isotopic purity with a rate of similar to 1000 kg/year of U-233 or Pu-239 can be produced. Radiation damage through atomic displacements and helium gas production after a plant operation period of 30 years is very low, namely dpa <1 and He < 2 ppm, respectively.Öğe Fissile fuel breeding with peaceful nuclear explosives(ELSEVIER SCIENCE SA, 2003) Sahin, S; Yalcin, S; Yildiz, KNeutron 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 (ThF4 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 U-233 or 3000 kg/year Pu-239 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, (C) 2003 Elsevier B.V. All rights reserved.Öğe Investigation of CANDU reactors as a thorium burner(Pergamon-Elsevier Science Ltd, 2006) Sahin, S; Yildiz, K; Sahin, HM; Acir, ALarge quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium can be used as a booster fissile fuel material in the form of mixed ThO2/PuO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. The paper investigates the prospects of exploiting the rich world thorium reserves in CANDU reactors. Two different fuel compositions have been selected for investigations: (1) 96% thoria (ThO2) + 4% PuO2 and (2) 91% ThO2 + 5% UO2 + 4% PuO2. The latter is used for the purpose of denaturing the new U-233 fuel with U-238. The behavior of the reactor criticality k(infinity) and the burn-up values of the reactor have been pursued by full power operation for >similar to 8 years. The reactor starts with k(infinity) = -1.39 and decreases asymptotically to values of k(infinity) > 1.06, which is still tolerable and useable in a CANDU reactor. The reactor criticality k(infinity) remains nearly constant between the 4th year and the 7th year of plant operation, and then, a slight increase is observed thereafter, along with a continuous depletion of the thorium fuel. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn-up can be achieved with the same fuel (> 160,000 MW D/MT). The reactor criticality would be sufficient until a great fraction of the thorium fuel is burned up, provided that the fuel rods could be fabricated to withstand such high burn-up levels. Fuel fabrication costs and nuclear waste mass for final disposal per unit energy could be reduced drastically. (c) 2005 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 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 Power flattening in the fuel bundle of a CANDU reactor(ELSEVIER SCIENCE SA, 2004) Sahin, S; Yildiz, K; Acir, AThe strong non-uniformity of the fission power production density in the CANDU fuel bundle could have been mitigated to a great degree. A satisfactory power flattening has been achieved through an appropriately evaluated method by varying the composition of the LWR spent fuel/ThO2 Mixture in a CANDU fuel bundle in radial direction and keeping fuel rod dimensions unchanged. This will help also to greatly simplify fuel rod fabrication and allow a higher degree of quality assurance standardization. Three different bundle fuel charges are investigated: (1) the reference case, uniformly fueled with natural UO2, (2) a bundle uniformly fueled with LWR spent fuel, and (3) a bundle fueled with variable mixed fuel composition in radial direction leading to a flat power profile (100% LWR spent fuel in the central rod, 80% LWR + 20% ThO2 in the second row, 60% LWR + 40% ThO2 in the third row and finally 40% LWR + 60% ThO2 in the peripheral fourth row). Burn-up grades for these three different bundle types are calculated as similar to7700, similar to27,300, and 10,000 MW.D/MT until reaching a lowest bundle criticality limit of k(infinity) = 1.06. The corresponding plant operation periods are 170, 660, and 240 days, respectively. (C) 2004 Elsevier B.V. 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 Radiation shielding mass saving for the magnet coils of the VISTA spacecraft(PERGAMON-ELSEVIER SCIENCE LTD, 1999) Sahin, S; Sahin, HMRadiation shielding structure of a design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit has been investigated. Fusion power deposited in the inertial confined fuel pellet debris delivers the rocket propulsion with the help of a magnetic nozzle. The nuclear heating in the super conducting magnet coils determines the radiation shielding mass of the spacecraft. It was possible to achieve considerable mass saving with respect to a recent design work, coupled with higher design limits for coil heating (up to 5 mW/cm(3)). The neutron and gamma-ray penetration into the coils is calculated using the SN methods with a high angular resolution in (r-z) geometry in S16P3 approximation by dividing the solid space angle in 160 sectors. Total peak nuclear heat generation density in the coils is calculated as 3.143 mW/cm(3) by a fusion power of 17 500 MW. Peak neutron heating density is 1.469 mW/cm(3) and peak gamma-ray heating density is 1.674 mW/cm(3). However, volume averaged heat generation in the coils is much lower, namely 74, 163 and 337 mu W/cm(3) for neutron, gamma-ray and total nuclear heating, respectively. The net mass of the radiation shielding for the magnet coils is 200 tonne by a total mass of 6000 tonne of the space craft. (C) 1999 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.
