Hacı Mehmet Şahin

Neutronics analysis of HYLIFE-II blanket for fissile fuel breeding in an inertial fusion energy reactor

A protective,60 cm thick flowing liquid wall coolant is investigated as energy carrier,and fusile and fissile breeder medium in an inertial fusion energy (IFE) reactor. Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF4 and UF4) starting with 2 mol% up to 12 mol%. For a plant operation period of 30 years,radiation damage values were found as DPA= 65 for 2 mol% heavy metal in the coolant,and remain practically constant with increasing heavy metal fraction,well below the presumable limit of DPA=100. Helium production values are calculated as 270 appm for 2 mol% heavy metal fraction,also being far below the limit value of 500 appm and remain at the same level with increasing heavy metal fraction. Such a flowing protective liquid wall extents the lifetime of the rigid first wall structure to a plant lifetime of 30 years. Fissionable metal salt in the flowing liquid enables one to breed high quality fissile fuel for external reactors by a self-sustaining tritium breeding for the fusion plant and increases plant power output. # 2002 Elsevier Science Ltd. All rights reserved.

Power flattening and minor actinide burning in a thorium fusion breeder

Abstract A neutronic analysis has been performed for a thorium fusion breeder with a special task of burning minor actinides 237 Np , 241 Am , 243 Am and 244 Cm and production of 233 U , 238 Pu , 242 m Am and 245 Cm for spacecraft application. 238 Pu is an important radioisotopic energy source for spacecraft generators. As potential nuclear fuels in the foreseeable future, 233 U , 242 m Am and 245 Cm 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 ThO 2 . 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 233 U production is ∼300 kg per metre of blanket height. The 238 Pu 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 245 Cm yield is 1.1 and 2 kg, respectively. The net 242 m Am 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.

Neutronic investigation of a hybrid version of the ARIES-RS fusion reactor

Abstract A hybrid version of the ARIES-RS of the commercial 1000 MW el power plant design concepts is investigated. A 10 cm fission zone at the inner blanket leads to a blanket multiplication of M =1.946 with ThC fuel or M =3.03 with UC fuel and increasing the fusion power from 2170 to 4200 MW or to 6500 MW, respectively. Despite a partial replacement of the lithium zone by the fissile zone, tritium breeding remains still >1.05, which will be required for a self-sustaining fusion driver. In addition to fusion power amplification, substantial fissile material will be produced at start-up conditions with a fission breeding rate of 233 U=0.183 (with ThC) or 239 Pu=0.263 (with UC) per incident fusion neutron, which correspond to 4410 kg 233 U/year or 6500 kg 239 Pu/year, respectively, by a full fusion power of 2170 MW. Damage calculations are based on the neutron flux load of 3.7 MW/m 2 and 5.6 MW/m 2 and have resulted with DPA=31 and 78 at the inner and outer first wall, respectively. DPA limit on vanadium will then require a change of the first walls at the inner and outer blankets after ∼6.5 and ∼2.5 full power years, respectively. Helium production at the inner and outer first wall is calculated as 117 and 237 ppm, respectively. This would lead to a first wall change after ∼4 and ∼2 full power years at the inner and outer blankets, respectively. The spectrum softening in the fissile zone will cause a relatively lower material damage as compared with the pure fusion reactor design.

Investigation of the effects of the resonance absorption in a fusion breeder blanket

Abstract In generic neutronic studies of a fusion breeder, resonance absorption is mostly neglected. In this work, the effects of resonances on the neutronic parameters a fast and moderated fusion blanket are investigated. In the present case, a (D,T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing 10 rows in radial direction, covers the cylindrical fusion plasma chamber. The fissile fuel is natural UO2. Fissile zone is cooled (a) with pressurised helium gas for the fast blanket and (b) with light water for the moderated, each of them with a volume ratio of Vcoolant/Vfuel=2 in the fissile zone. The study has shown that careful resonance self-shielding calculations are indispensable for neutronic studies of a moderated fusion blanket. On the other hand the omission of the resonance self-shielding in generic studies of a fast fusion blanket can be tolerated to some degree. Furthermore, a correct description of the fusion neutron source spectrum has great importance on neutronic parameters, along with the resonance self-shielding calculations.

Effects of spectral shifting in an inertial confinement fusion system

Abstract The 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 6Li2DT 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 ∼ 1000 kg/year of 233U or 239Pu 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.

Minor actinide burning in a CANDU thorium reactor

Abstract Nuclear waste actinides can be used as a booster fissile fuel material in form of mixed fuel with thorium in a CANDU reactor in order to assure the initial criticality at startup. Two different fuel compositions have been found useful to provide sufficient reactor criticality over a long operation period: 1) 95% thoria (ThO2)+5% minor actinides MAO2 and 2) 90% ThO2+5% MAO2+5% UO2. The latter allows a higher degree of nuclear safeguarding through denaturing the new 233U fuel with 238U. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 10 years. The criticality starts by k∞>1.3 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2nd year and remains close to k∞ = ∼1.06 for ∼10 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn up could be achieved with the same fuel material (up to 200000 MW.D/MT), provided that the fuel rod claddings would be replaced periodically (after every 50000 or 100000 MW.D/MT). The reactor criticality can be maintained until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.

Performance analysis of 233U for fixed bed nuclear reactors

Abstract Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigated for the mixed fuel 233UO2/ThO2 as an alternative to low enriched 235UO2 fuel. CERMET fuel with a zirconium matrix and cladding has been used throughout the study. The main results of the study can be summarized as follows: Reactor criticality is already achieved by ∼2% 233UO2 with the mixed 233UO2/ThO2 fuel. At higher 233U fractions, reactor criticality rises rapidly and exceeds keff > 1.5 already by 9% 233UO2. With 100% 233UO2, start up criticality can reach keff = 2.0975. Time dependent reactor criticality keff and fuel burn up have been investigated for two different mixed fuel 233UO2/ThO2 compositions, namely: 4% 233UO2 + 96% ThO2 for a reactor power of 40 MWel (120 MWth) and 9% 233UO2 + 91% ThO2 for a reactor power of 70 MWel (210 MWth). Sufficient reactor criticality (keff > 1.06) for continuous operation without fuel change can be sustained during ∼5 and 12 years with 4% and 9% 233UO2 fractions in the mixed fuel, leading to burn ups of ∼36000 and >105000MWD/t, respectively. Thorium based fuel produces no prolific uranium. Plutonium production remains negligible.

Optimization of the radiation shielding mass for the magnet coils of the VISTA spacecraft

An attempt has been made for the optimisation of the radiation shielding of a spacecraft design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris, and with the help of a magnetic nozzle. The allowable nuclear heating in the super conducting magnet coils (up to 5 mW/cm3) is the crucial criterion for the dimensioning of the radiation shielding structure of the spacecraft. The optimized design reduced the shield mass from 600 tons to 93 and 88 tons with natural and enriched lithium, respectively. The space craft mass was 6000 tons. Total peak nuclear power density in the coils is calculated to be 5.0 mW/cm3 for a fusion power of 17,500 MW. Peak neutron heating density is 2.6 mW/cm3 and peak γ-ray heating density is 2.9 mW/cm3 (all on different points). However, volume averaged heat generation in the coils is much lower, namely 0.30, 0.73 and 1.03 mW/cm3 for neutron, γ-ray and total nuclear heating, respectively.

1.20 Nuclear Energy

This chapter presents a general overview on nuclear fission energy production. Fundamentals of the fission process and fission energy production are highlighted. The conventional and advanced nuclear fuels for fission reactors, fusion reactor fuels, high temperature coolants, and their properties are outlined. The study covers conventional and advanced nuclear reactor types, generation-IV reactor, space nuclear reactors, and floating reactors. Environmental issues are also addressed. Finally, the sustainability and renewability potential of nuclear energy is demonstrated with CANada Deuterium-Uranium Reactor reactors, as well as with the introduction of high temperature reactors and fusion reactors into the energy sector.

Utilization Potential of Thorium in CANDU Reactors and in Fusion–Fission (Hybrid) Reactors

World thorium reserves are approximately three times more abundant than the natural uranium reserves. Turkey is rich is thorium resources. Hence, thorium remains a potential energy source for future energy strategies in Turkey.