A Thermal-hydraulic conceptual design of the energy production blanket for a fusion-fission hybrid reactor is proposed in this paper. The blanket uses U-10Zr alloy as its fuel, in which the uranium abundance is natural. It uses water as coolant as well as moderator, so that the Pu-239 bred from neutron capture reaction of U-238 can be depleted by thermal neutrons, while the fast neutrons can cause U-238 fission. A properly designed blanket needs no separation of Pu-239 from spent fuel, and makes full use of U-238 to produce energy. Based on the preliminary neutronic conceptual design of the blanket, a detailed blanket model has been established, and the distribution of deposited energy in it is calculated by MCNP. In this conceptual model, the coolant flows in tubes, which go through from the bottom to the top of the blanket, and the fuel fills in the gap between tubes. A set of criteria have been developed for the thermal-hydraulic design, consequently the restrictions on tube size, wall thickness, pitches and power density are determined. The study on thermal-hydraulic characteristics of the blanket is carried out by the 3D computational fluid dynamic code FLUENT, in which the variation of transverse cross section due to the special geometry of Tokamak can be considered. The temperature and heat flux distribution at typical positions are analyzed, and the dependence of thermal-hydraulic performance on fuel-coolant volume ratio is discussed. It can conclude that the conceptual design proposed in this paper has satisfactory performances, all parameters satisfy the preset thermal-hydraulic criteria, and have large enough safety margin.
- Nuclear Engineering Division
Thermal-Hydraulic Conceptual Design of Water-Cooled Blanket of a Fusion-Fission Hybrid Reactor for Energy Production
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Guo, H, Peng, X, Shi, X, & Ma, J. "Thermal-Hydraulic Conceptual Design of Water-Cooled Blanket of a Fusion-Fission Hybrid Reactor for Energy Production." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors. Chengdu, China. July 29–August 2, 2013. V002T05A008. ASME. https://doi.org/10.1115/ICONE21-15416
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