Neutronic analysis of a moderated (D, T) fusion driven hybrid blanket


Yapici H. , Akansu S. O. , Ozceyhan V.

ANNALS OF NUCLEAR ENERGY, cilt.27, ss.1237-1244, 2000 (SCI İndekslerine Giren Dergi) identifier identifier

  • Cilt numarası: 27 Konu: 13
  • Basım Tarihi: 2000
  • Doi Numarası: 10.1016/s0306-4549(99)00122-x
  • Dergi Adı: ANNALS OF NUCLEAR ENERGY
  • Sayfa Sayıları: ss.1237-1244

Özet

In this study, neutronic effects on (D-T) driven hybrid reactor fuelled with different mixtures of 45.5% fertile fuels, vary from 0 to 100% for uranium components and from 100 to 0% for the thorium component in the fuel mixtures. These mixtures are UF4 + THF4, UO2 + ThO2 UC2 + ThC2. The coolants are selected as flibe (Li2BeF4), natural lithium and air for the heat transfer out of the fuels zones and the clad material is selected as SS-316 stainless steel with a volume fraction ratio of 9%. The fuel zones of the hybrid reactor are investigated to obtain fissile fuel breeding (U-233 and/or Pu-239). A regeneration period of up to 48 months is investigated by a plant factor of 75% under a first wall neutron flux (phi(w)) of 2.2210(14) (14.1 MeV) n/ cm(2).s for a conventional (D,T) driven hybrid reactor. These correspond to a first wall neutron load of 5 MW/m(2). At the end of the operation periods, cumulative fissile fuel enrichment (CFFE) value varies between 6.7% in pure UF4 with flibe coolant and 3.5% in pure ThO2 with natural lithium coolant. At the beginning of the operation, the tritium breeding ratio (TBR) is higher than unity in air coolant and natural lithium coolant cases. In flibe coolant cases the TBR value can reach nearly 55% uranium component + 45% thorium components. Therefore, the hybrid blanket is self-sufficient with respect to the TBR for these cases. It is important to follow the non-prolific level of the plutonium fuel during operation because of the nuclear weapons hazard. Nuclear quality of the plutonium increases linearly during the operation period. Pu-240 content has to be higher than 5% for safety. Pu-240 content is higher than 5% in UF4 + ThF4, UC2, + ThC2 and (UO2 > 20%) + (ThO2 < 80%) with flibe coolant. This is very important criteria for safety. In natural lithium and air coolant cases Pu-240 content is lower than 5%. In these cases, the operation period must be increased for safety. U-233 is used as a fuel in thermal reactors and can be separated from the Thourium component. But this separation is too difficult in the thorium component + uranium component mixtures. Therefore, investigated hybrid blanket is safer than 100% Th component. (C) 2000 Elsevier Science Ltd. All rights reserved.