Neutronic Study Of Thorium Fueled Blanket For Fusion-fission Hybrid Energy Reactor

The natural thorium fuel inventory on earth is about three times than the uranium. Fusion-fission hybrid reactor(FFHR) could efficiently convert 232 Th into fissile fuel 233 U using its abundant fusion neutron source. The fissile nuclear fuel inventory could be greatly enlarged utilizing FFHR. What’s more, the requirement on the fusion plasma parameter in the FFHR is much lower than the pure fusion reactor due to the energy multiplication capability of its blanket. Tokamak parameters of International Thermonuclear Experiment Reactor(ITER) which is under construction nowadays are enough for the FFHR to use the fusion power in advance. One problem with the utilization of thorium in FFHR is the low energy multiplication factor(M) of blanket due to the poor neutronic performance of 232 Th. It will bring higher requirement on the fusion power and fusion gain which is not desirable. In order to make the FFHR more competitive, only natural thorium and uranium fuel are used. The reactor aims to maximize the 233 U breeding rate under the requirement of high M(M≥6.0) and tritium breeding rate(TBR≥1.05) in the reactor’s entire life. In addition, the bred 233 U could be burnt in site with closed Th-U fuel cycle.Conceptual design of the blanket with above purposes is carried out. The code system COUPLE2.0, developed by the Institute of Nuclear and New Energy Technology, Tsinghua University, is used to simulate the nucleus’ s depletion process of the FFHR blanket. The main content of this thesis is in the following:1. A simplified two dimensional “D” shaped blanket, which takes the fusion core parameter of ITER as reference, is constructed in this paper. An innovative design for a water cooled FFHR, aiming at efficiently utilizing natural uranium and thorium resources, is presented. The fission blanket is designed with two types of modules, i.e. the natural uranium modules(U-Modules) and thorium modules(Th-Modules), which are alternately arranged in the toroidal and poloidal directions of the blanket. 60% volume fraction in the blanket’s toroidal direction could be inserted with thorium module in the first core. After the first core operates 20 years, the uranium module is then replaced by new thorium fuel module. The all thorium fueled second core could burn up about 90 tonnes 232 Th at the end of 60 years operating.2. A three dimensional water cooled blanket with thorium and uranium module arranged alternately in the poloidal direction is constructed. M and TBR of 2-D model are 10% larger than the 3-D model’s results. The 3-D blanket is optimized to reach the maximum 233 U breeding rate with requirement of M and TBR. The 3-D blanket could also transit to all thorium fueled mode after 20 years operating of the first core. In order to improve the 233 U breeding rate and decrease the transition time between first and second core, a fission-suppressed thorium module, helium cooled thorium module and molten salt cooled fast fission blanket are designed. The 233 U breeding rates of above three thorium modules are 2, 2.5 and 3.2 times as much as the water cooled thorium module respectively.3. What’s more, a thorium bearing molten salt blanket with 233 U as the start fuel is designed. Energy multiplication and tritium breeding are accomplished in different regions. With(ThF4+233UF4) mole fraction set as 4.5% and the mole fraction of 233UF4 in(ThF4+233UF4) set as 12% in energy generation region, 6Li enrichment set as 15% in tritium breeding region, the simulated results shows, with on line periodic fuel reprocessing, with 8.2 tons 233 U start fuel, the blanket could operate with excellent performance, M(8~11), TBR(1.01~1.12) continuously and only natural thorium fuel is added into the system.