Sümer Şahin

Preliminary design studies of a cylindrical experimental hybrid blanket with deuterium-tritium driver

AbstractThe AYMAN research project has been initiated to formulate the main structure of a prototypical experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. This geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect to neutronic considerations.In this project, the fusion chamber is simulated by a cavity with a diameter of ∼1.6 m inside a cylindrical blanket. Fusion neutrons of 14 MeV are produced by a movable target along the axis of the cylinder. The movable neutron source allows simulation of a line source for integral experiments, which is a result of the linear nature of the Boltzmann transport equation.The calculations have shown that a blanket with a 13-cm-thick natural UO2 fuel zone and a 17-cm-thick Li2O zone has a self-sustaining tritium breeding for the fusion driver. By an appropriate dispersion of the Li2O zone inside the graphite reflector, it became possible to decrease the neutron leakage out of the reflector...

Potential of a Catalyzed Fusion-Driven Hybrid Reactor for the Regeneration of Candu Spent Fuel

In this paper the potential of a catalyzed fusion-driven fast hybrid blanket to regenerate Canada deuterium uranium (CANDU) spent fuel is investigated. The investigations are done to achieve enrichment grades of fissile isotopes (EGFIs) in four applications: recycling in a conventional commercial CANDU reactor (EGFI = 0.71 to 0.9%); recycling in an advanced conceptual CANDU reactor with a high burnup rate (EGFI = 1%); recycling in an advanced breeder with thorium fuel (EGFI {gt} 1.5%); recycling in a conventional light water reactor (LWR) (EGFI {gt} 3%). The regeneration periods of 5 to 7, 6 to 9, 12 to 15 and {gt}30 months, respectively, are evaluated for the four reactor types under a first-wall fusion neutron current load of 10{sup 14} (14.1-MeV n)/cm{sup 2} {center dot} s, corresponding to 2.64 MW/m{sup 2} and a plant factor of 75%. During the regeneration process, the burnup rates vary from 2000 MWd/t (for conventional CANDU) to 10,000 MWd/t (for LWRs), so that multiple recycling becomes possible.

Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation

The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO2; (2) ThO2 denaturated with 10% natural-UO2 and (3) ThO2 denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li2BeF4’) are selected for the nuclear heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m2 and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO2 produces weapon grade 233U. The denaturation of the 233U fuel is realized with a homogenous mixture of 90% ThO2 with 10% natural-UO2 as well as with 10% LWR spent nuclear fuel. The homogenous mixture of 90% ThO2 with 10% natural-UO2 can successfully denaturate 233U with 238U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO2 with 10% LWR spent nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.

Investigation of the Neutronic Potential of Moderated and Fast (D,T) Hybrid Blankets for Rejuvenation of CANDU Spent Fuel

AbstractThe potential of moderated and fast hybrid blankets is investigated for the rejuvenation of CANDU spent fuel. The moderated hybrid blanket has revealed poor neutronic performance and is not suitable for CANDU spent-fuel rejuvenation. The fast-fissioning hybrid blanket has excellent neutronic performance and is investigated to achieve different enrichment grades of fissile isotopes (EGFI) for three different applications:1. recycling in a conventional commercial CANDU reactor (EGFI = 0.71 to 0.9%)2. recycling in an advanced CANDU reactor concept with high burnup rate (EGFI = 1%)3. recycling in an advanced breeder with thorium fuel (EGFI > 1.5%)For fast-fissioning blankets, a rejuvenation period of up to 24 months is investigated by a plant factor of 75% under a first-wall (deuterium, tritium) fusion neutron current load of 1014 to 14 MeV·n/cm2·s, corresponding to 2.25 MW/m2.Rejuvenation periods of 6 to 12 months, 10 to 16 months, and 18 months, respectively, are evaluated for the above-mentioned re...

Proliferation hardening and power flattening of a thorium fusion breeder with triple mixed oxide fuel

Abstract The proliferation hardening of the 233 U fuel in a thorium fusion breeder has been realised successfully with a homogenous mixture of ThO 2 , natural-UO 2 and CANDU spent nuclear fuel in the form of a triple mixed oxide (TMOX) fuel. The new 233 U component will be successfully hardened against proliferation with the help of the 238 U component in the natural-UO 2 and spent fuel. The plutonium component remains non-prolific through the presence of the 240 Pu 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 ThO 2 component in the rods continuously in the radial direction (from 91 down to 64%). Three different coolants (pressurised helium, natural lithium and Li 17 Pb 83 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 232 Th (ThO 2 ), in the 238 U (natural-UO 2 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 ( 233 U+ 235 U+ 239 Pu+ 241 Pu) in the TMOX 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 TMOX fuel can achieve an enrichment degree of ∼1% after ∼12–15 months. A longer irradiation period (∼ 30 months) increases the fissile fuel enrichment levels of the TMOX towards much higher degrees (∼ 2%), 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.

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.

The Effect of the Spontaneous Fission of Plutonium-240 on the Energy Release in a Nuclear Explosive

Assuming the spontaneous fission neutron level as a neutron source, and using point kinetic methods in the course of the analytical treatment, the energy excursion of hypothetical nuclear explosives with mixed plutonium of various isotope compositions has been investigated. In this work various plutonium mixtures are compared with weapons-grade plutonium, regarding the influence of the spontaneous fission on the energy excursion in an atomic bomb. 17 refs.

Spent mixed oxide fuel rejuvenation in fusion breeders

Abstract A 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 238 U. 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 ( 235 U+ 239 Pu+ 241 Pu) 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 ∼7 months, suitable for reutilization in a CANDU reactor. LWR spent fuel requires >15 months to reach an enrichment grade ∼3.5%, suitable for reutilization in a LWR. A longer rejuvenation period (up to 48 months) increases the fissile fuel enrichment levels 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.

Neutronic Investigation of a Power Plant Using Peaceful Nuclear Explosives

A neutron physics analysis of the modified PACER concept was conducted to assess the required liquid zone thickness of which the volume fraction is 25% in the form of Li{sub 2}BeF{sub 4} (Flibe) jets and 75% as void. These liquid jets surround a low-yield nuclear fusion explosive and protect the chamber walls. The neutronic calculations assumed a 30-m-radius underground spherical geometry cavity with a 1-cm-thick stainless steel liner attached to the excavated rock wall. Achievement of tritium breeding ratios of 1.05 and 1.15 requires a Flibe thickness of 1.6 and 2.0 m, respectively, which results in average energy densities of 24900 and 19085 J/g. The authors` calculations show that for a Flibe zone thickness >2.5m, the activation of the steel liner and rock would be low enough after 30 yr of operation that the cavity would satisfy the U.S. Nuclear Regulatory Commission`s rules for {open_quotes}shallow burial{close_quotes} upon decommissioning, assuming other sources of radioactivity could be removed or qualified as well. This means that upon decommissioning, the site could essentially be abandoned, or the cavity could be used as a shallow burial site for other qualified materials. 25 refs., 5 figs., 8 tabs.

Fission power flattening in hybrid blankets using mixed fuel

In a source-driven fissionable blanket, a flat fission power density (FPD) is achieved by using a mixed fuel (ThO/sub 2/ and natural UO/sub 2/) with the thoriumuranium ratio changing from front to back in the ten fuel rows along the radial direction. A straightforward graphic method is used. The temporal behavior of the FPD has been observed for an operation period of 6 months and for a plant load factor of 75% by applying a fusion driver neutron flux of 10/sup 14/ 14-MeV neutrons(cm/sup 2/ . s) at the first wall, corresponding to --2.25 MWm/sup 2/. To keep the power density flat, it is necessary to replace the fuel in rows 1, 2, and 3, close to the first wall. The time intervals for this operation increase, counting from initial start-up, typically, 2 months, 6 months, etc. One result of this study is that plutonium produced in such a hybrid blanket contains very low amounts of even isotopic components even over very long operation times of --3 yr. Hence, if fusion reactors are introduced into the energy market, special regulations are needed for international safeguarding.