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Neutronic Study of a Molten Salt Cooled Natural Thorium–Uranium Fueled Fusion–Fission Hybrid Energy System

Research paper by S. C. Xiao, J. Zhao, Z. Zhou, Y. Yang

Indexed on: 19 Nov '14Published on: 19 Nov '14Published in: Journal of Fusion Energy



Abstract

In this paper, a preliminary study on the neutron physics characteristics of a molten salt cooled fast fission blanket for a new type fusion–fission hybrid reactor (FFHR) aiming at efficiently utilizing the natural thorium resource and electric power generation is presented. The major objective is to study the feasibility of this fast fission concept with multi-purposes, including energy gain, tritium breeding ratio (TBR) and 233U breeding rate. In order to improve overall neutron economy of the system, the blanket adopts the seed-blanket concept and consists of two main kinds of modules, i.e. the natural uranium fuel module (U-module) as the seed and thorium fuel module (Th-module) as the blanket. The uranium module plays the dominate role in the energy production and neutron multiplication. Excess neutrons produced by the uranium modules are then used to breed 233U fuel and tritium. The COUPLE2 code developed by the Institute of Nuclear and New Energy Technology of Tsinghua University is used to simulate the neutronic behaviour in the blanket. The simulated results show that with 505 tons thorium fuel loading, system multi-purpose, i.e. moderate energy multiplication (initial M ≥6), tritium self sufficiency and high 233U breeding rate, could be reached simultaneously. The preliminary results indicate that it is rather promising to design a high-performance molten salt cooled fission blanket of FFHR for electric power generation and 233U breeding by directly loading natural uranium and thorium if an ITER-scale 500 MW tokamak fusion neutron source is achievable.