Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

• Main Authors: Zhu Guifeng, Zou Yang, Xu Hongjie
• Format: Article
• Language: English
• Published: EDP Sciences 2016-01-01
• Series: EPJ Nuclear Sciences & Technologies
• Online Access: http://dx.doi.org/10.1051/epjn/e2015-50032-6

Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern) and two kinds of reflector materials (SiC and graphite). This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9Be(n,2n) reaction and low neutron absorption of 6Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.