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

Akansu S. O. , Unalan S.

ANNALS OF NUCLEAR ENERGY, vol.29, no.3, pp.287-302, 2002 (Journal Indexed in SCI) identifier identifier

  • Publication Type: Article / Article
  • Volume: 29 Issue: 3
  • Publication Date: 2002
  • Doi Number: 10.1016/s0306-4549(01)00038-x
  • Title of Journal : ANNALS OF NUCLEAR ENERGY
  • Page Numbers: pp.287-302


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 (Li2BeF4) and fueled by UO2 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 approximate to8.5, approximate to9.3, approximate to8.4 and approximate to8.2% for UO2 with LWR rods, LWR spent fuel, UO2 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 Pu-240 in the produced plutonium is higher than 7%. However, other cases with UO2 fuel can reach sufficient safety after an operation period of 30 months. (C) 2001 Elsevier Science Ltd. All rights reserved.