Hüseyin Yapıcı

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...

Neutronic performance of proliferation hardened thorium fusion breeders

Abstract Production 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 ∼4%, 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 238U component of natural-UO2 can provide a limited proliferation hardening only for the 233U 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.

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.

Investigation of neutronic potential of a moderated (D–T) fusion driven hybrid reactor fueled with thorium to breed fissile fuel for LWRs

Abstract The potential of a moderated hybrid reactor fueled with ThC 2 or ThF 4 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 ThC 2 or ThF 4 are replaced in the fissile fuel zone of the hybrid reactor. It is considered that gas (He or CO 2 ), or flibe (Li 2 BeF 4 ), 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 ThF 4 fuel and flibe coolant mode, and the lowest CFFE value 3.55% is in the ThC 2 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 ∼25%. The lowest radial neutron leakage out of the blanket is in the blanket with the flibe coolant modes.

Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO2, ThC, UO2, UC, U3Si2 and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of 239Pu breeding capability among the uranium fuels whereas UO2 has the lowest one. 239Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and 233U production values are 0.125 and 0.140 in the blanket fueled with ThO2 and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm3 in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.

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.

Numerical solutions of conjugate heat transfer and thermal stresses in a circular pipe externally heated with non-uniform heat flux

Conjugate heat transfer by forced convection through an externally heated pipe has many important engineering applications. In the present work, the radial and axial heat conductions and thermal stresses in a pipe with uniform or non-uniform wall heat flux of fully developed laminar forced convective conjugate heat transfer have been considered for analysis. The analysis is based on the two-dimensional steady-state heat conduction equation and laminar boundary layer equation for the flowing fluid by using a finite difference scheme. Water has been used as a fluid. Numerical calculations have been performed by using the FLUENT 4.5 and HEATING7 computer codes. The temperature and stress ratio distributions inside the pipe wall, heated from the outer surface by applying uniform and non-uniform heat fluxes, have been presented for two different mean flow velocities. The temperature distributions of the flowing fluid inside the pipe have also been presented for all investigated cases.

Study on transmutation of minor actinides discharged from high burn-up PWR-MOX spent fuel in the force-free helical reactor

Abstract The force-free helical reactor (FFHR) is a demo relevant helical-type D-T fusion reactor. In this study, the burning and/or transmutation (B/T) of minor actinides (MAs) has been investigated for an operation period (OP) of up to 10 years in the FFHR by 75% plant factor (η) under a neutron wall load (P) of 1.5 MW/m2. In order to incinerate and transmute the MAs effectively, transmutation zone (TZ), containing the mixture of MA nuclides discharged from high burn-up pressured water reactor (PWR)-MOX spent fuel, has been located in the blanket of the FFHR. The MA mixture has been spherically prepared, and cladded with SiC to prevent fission products from contaminating coolant and the MAs from contacting coolant. Helium has been selected for the nuclear heat transfer in the TZ. Effect of the MA volume fraction in the zone on the B/T has been also investigated. The energy multiplication ratio (M), which is one of the main parameters in a fusion-fission hybrid reactor and relates to fission rate (RF), is quite high and increases from 5.8 to 9.4 in the case of a MA fraction of 10% and from 15.3 to 31.8 in that of 20% depending on the OP. The neutron multiplication coefficient (keff), also relating to the RF, is less than 0.9 in all investaged cases during the OP. Tritium breeding ratio (TBR) value is greater than 1.1 for all investigated cases so that tritium self-sufficiency is maintained for (D,T) fusion driver. Its value reaches to 3.2 in the case of a MA fraction of 20% at the end of operation period (EOP). The spatial non-uniformity of the fission energy density can be expressed with the help of the peak-to-average fission power density ratio (Γ). The Γ value varies in the range of 1.05–1.14 depending on the MA fraction and the OP. These values show that the TZ has a good flat fission power density profile in all investigated cases during the OP. At the beginning of operation period (BOP), the total transmutation rate (TR) values of 237Np, 241Am and 243Am are 1489 and 1206 kg/GWthyr in the cases of 10 and 20% MA, respectively. The TRs of 237Np, 241Am and 243Am decrease exponentially in all investigated cases during the OP. At the EOP, in the case of a MA fraction of 20%, the effective half-lives of 237Np, 241Am and 243Am decrease to 5.4, 5.1 and 6.8 years, repectively, and the net transmutation fractions (TFs) for the whole TRUs and MAs are obtained as 40 and 60%, respectively. These consequences bring out that the blanket carries out the B/T of MA effectively.