Sebahattin Ünalan

Rejuvenation of the CANDU spent fuel in (D-T)-driven hybrid reactors

The possibility of Canada deuterium uranium reactor (CANDU) spent-fuel rejuvenation in deuterium-tritium (D-T)-driven hybrid reactors having 17.8 cm of fissile zone thickness is investigated for various plasma chamber dimensions (DR = 18.7, 118.7, 218.7, and 418.7 cm) with a linear fusion neutron source (plasma dimension is assumed as DR/2) under different first-wall loads (P w = 2, 4, 6, 8, and 10 MW/m 2 ). The behavior of the spent fuel is observed during 36 months for discrete time intervals of At = 15 days and by a plant factor of 75%. The fissile fuel zone is considered to be cooled with three different coolants: gas (He or CO 2 ), Flibe (Li 2 BeF 4 ), and natural Li. As a result of the calculation, in the case ofthe first-wall load and the plasma chamber dimensions being selected high, although the first-wall material had been damaged considerably by the high neutron flux (displacements per atom > 100 and He > 500 parts per million for P w > 2 MW/m 2 over 3 yr of operation) and the maximum temperature in the centerline of the fuel rod (T m ) had reached the melting point (T m > 2600°C for P w > 6 MW/m 2 and DR > 1 m), it was observed that the neutronic performance of the hybrid reactor improved unnegligibly. For DR = 18.7 cm, at the beginning of rejuvenation, the tritium breeding ratio values were 1.18, 0.85, and 1.26 for gas, Flibe, and natural Li, respectively, and by the end of the rejuvenation had increased to 1.26, 0.93, and 1.32 for 2 MW/m 2 and to 1.60, 1.28, and 1.58 for 10 MW/m 2 . In addition, the blanket energy multiplication M increased to 5.47, 4.92, and 5.02 for 2 MW/m 2 and to 8.89, 9.32, and 7.58 for 10 MW/m 2 from 4.64, 3.90, and 4.37, respectively. Only for Flibe, when the DR value is preferred at ∼1 m, the M values increased to 5.82 and to 15.0 from 3.92 for 2 and 10 MW/m 2 , respectively. Under the same conditions, the average cumulative fissile fuel enrichment (CFFE) values indicating the rejuvenation performance increased to 1.67, 2.13, and 1.57% for 2 MW/m 2 and to 5.78, 7.69, and 5.39% for 10 MW/m 2 from 0.418%, respectively. For Flibe coolant, while the same CFFE value is 11.2% at DR = ∼1 m, it is 11.4% at DR = ∼4 m. The contribution of a large plasma chamber (DR > I m) to neutronic performance can be neglected. The best rejuvenation performance and neutron economy have been shown by Flibe. However, Flibe has shown a bad capability until the rejuvenation time in terms of tritium breeding. For Flibe, to breed enough tritium, the values of the first-wall load and required rejuvenation time must be larger than 6 MW/m 2 and 2 yr, respectively. For all cases, the denatured character of the initial fuel charge remains denatured for all investigated cases over the whole plant operation period in a hybrid reactor although the Pu quality increases continuously during the rejuvenation process. In addition, our calculations have proven that the effects of the important fission products ( 135 Xe, 149 Sm) and plasma densities up to 10 21 (D + T)/cm 3 can be neglected for the neutronic performance ofthe hybrid reactor, which rejuvenates CANDU spent fuel.

Investigation of the flattened fissile fuel enrichment possibility with a (D, T) driven hybrid blanket cooled by flibe (Li2BeF4)

Abstract This work investigated the possibility of the fuel production with flattened cumulative fissile fuel enrichment (CFFE) at the hybrid reactor, to be cooled by the flibe (Li 2 BeF 4 ) and fueled by UO 2 with a LWR fuel rod and CANDU fuel rod diameters, LWR spent fuel and CANDU spent fuel, with an original fuel rod diameter during the operation period of 5 years. For that purpose, the new fuel zone structure is provided by means of the ratio of the flibe to fuel per fuel row in the fuel zone being varied. Neutronic performance of the (D, T) driven hybrid blanket with this fuel zone is followed by a plant factor of 75% under a first wall load of 5 MW/m 2 . The fuel row numbers are selected as 10. For all fuels, the possibility of the fuel production having almost the same CFFE in all fuel rows of the fuel zone of the hybrid blanket is possible by a deviation of 2%. Moreover, the fissile fuel production capability of the suggested blanket increased considerably. However, tritium breeding ratios and the displacement-per-atoms (dpa) values in the first wall and clad material are almost not affected by this blanket structure, the energy production decreased slightly. At the end of the operation period of 5 years, the CFFE value reached ≈8.5, ≈9.3, ≈8.4 and ≈8.2% for UO 2 with LWR rods, LWR spent fuel, UO 2 with CANDU rods and CANDU spent fuel, respectively. The remaining fuel from hybrid blankets with CANDU and LWR spent fuels have enough safety from the viewpoint of the plutonium non-proliferation since the isotropic percentage of 240 Pu in the produced plutonium is higher than 7%. However, other cases with UO 2 fuel can reach sufficient safety after an operation period of 30 months.

Improvement of the Neutronic Performance of the Hybrid Reactor Rejuvenating Spent Fuels Using Various Moderators

Effects on the neutronic performance of the hybrid blanket rejuvenating light water reactor and CANDU spent fuels of moderators (Be, C, and D2O) inserted between the fusion chamber and the fissile ...

Improvement of the Inertial Fusion Energy Reactor Performance by Means of UF4 and ThF4

In an inertial fusion energy (IFE) reactor of 1000-MW(electric) fusion power, 95% flibe and 5% fuel with DRc thickness instead of 100% flibe are used. At startup, the tritium breeding ratio and M-blanket energy multiplication ratio are 1.05 and 1.26 for UF4 and DRc ≈ 60 cm, respectively. These values increase during an operation period of 30 yr. In 11 yr, M increases from 1.26 to 2 [= 2000 MW(electric)]. After operation of 11 yr, the energy production is stabilized by means of separation of produced plutonium. After 30 yr, displacement per atom (dpa) and helium production in the first wall are calculated as 92 dpa and 590 ppm, respectively. In addition, the cost of electricity values of the HYLIFE-II and the improved HYLIFE-II of 2000 MW(electric) drop from 4.5 and 3.2 ¢/kW⋅h to 4.18 and 3.00 ¢/kW⋅h, respectively. On the other hand, the IFE reactor has the fissile fuel breeding potential of 70 tonnes. The fissile fuel of 45 tonnes corresponding to ≈ 2350 kg/yr would be sufficient to provide makeup fuel for ≈ 10 light water reactors after 11 yr. After the shutdown process, 25 tonnes of fissile fuel with fuel enrichment of 23% would be left over.

Investigation of heat transfer and pressure drop for various obstacles in a rectangular‐sectioned 90° bend

Purpose – Re<95,000 based on hydraulic diameter, heat transfer and turbulent flow through a rectangular‐sectioned 90° bend was investigated numerically and experimentally. To develop turbulence level, square prism and cylindrical obstacles was placed in the center of the bend.Design/methodology/approach – For heat transfer, uniform heat flux of 5,000 W/m2 from bend surfaces is assumed. Numerical analysis was realized for both the turbulent flow and heat transfer. For numerical study, FLUENT 6.1.22 code, RSM turbulence model, hybrid hexahedral‐tetrahedral cell structures and uniform inlet velocity assumption were selected. For the pressure distribution in the bend and velocity profile at the outlet of the bend, the experiments was carried out by means of manometers with ethyl alcohol, Mano‐air 500 Equipment and pitot‐static tube.Findings – There was a high level of validation obtained between the numerical and the experimental results. Thereby, the mentioned numerical calculation method can be used most en...

Neutronic and Thermal Analysis of a Peaceful Nuclear Explosion Reactor

Thermal and neutronic behavior of a peaceful nuclear explosion reactor (PACER) producing ≈ 1.2 GWe electrical-power from fusion explosions in a cylindrical explosion chamber (radius = 30 m, height = 75 m) are analyzed. For determination of flibe mass (m) required for safe operation temperatures and pressures with enough tritium breeding ratio (TBR) and high M (fusion energy absorption ratio), neutronic calculations are carried out for different coolant zone positions (DR) and coolant zone thicknesses (DRc). Inlet pressure and temperatures (Tin) of flibe are 1 atm, and 823 and 1540 K. For all DR values, TBR and M reached saturation values of 1.27 and 1.07 at certain DRc values, respectively. Thereby, m increases with increased DR. To decrease flibe mass requirements, DR must be as low as possible. However, this causes high equilibrium pressures and enormous temperatures. Therefore, to decrease mechanical and chemical damages on the walls, DR must be high. The highest equilibrium pressures for the investigated parameters are ≈ 100 and ≈ 160 atm for Tin = 823 K and Tin = 1540 K, respectively. For the equilibrium temperature and pressures of 1750 K and ≈ 20 atm, m and DR should be 3000 tonnes and 400 cm for Tin = 823 K, and 25000 tonnes and 700 cm for Tin = 1540 K.