NDLI: Thermo-Hydraulic Analysis of the Optimized Helium Cooled Solid Breeder Blanket for CFETR

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Thermo-Hydraulic Analysis of the Optimized Helium Cooled Solid Breeder Blanket for CFETR

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Cui, Shijie Zhang, Dalin Chang, Jie Tian, Wenxi Qiu, Suizheng Su, G. H.
Copyright Year	2016
Abstract	China Fusion Engineering Test Reactor (CFETR) is under design recently, in which a conceptual structure of the helium-cooled solid breeder blanket is proposed as one of the candidate tritium breeding blankets. In this concept, three radial arranged U-shaped breeding zones are designed and optimized for higher Tritium Breeding Ratio (TBR) and structure simplification. This blanket uses the Li4SiO4 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized typical outboard blanket module are performed by CFD method, where the nuclear heating rate is obtained from the preliminary neutronics calculations. The thermal hydraulic behaviors of the first wall (FW), the temperature distributions of submodule structure material, Li4SiO4 pebble bed and Beryllium pebble bed under normal and critical conditions are calculated, respectively. The results show that the temperature on the blanket module can be effectively cooled below allowable temperature limits of the materials, even if the FW is suffering the maximum surface heat flux, which verified the reasonability of the design of the blanket cooling scheme. In addition, several parametric sensitivity studies are conducted to investigate the influences of main parameters (e.g. coolant mass flow rate, inlet temperature, pebble bed thermal conductivity and fusion power) on the temperature distributions of the blanket components.
Sponsorship	Nuclear Engineering Division
File Format	PDF
ISBN	9780791850022
DOI	10.1115/ICONE24-60144
Volume Number	Volume 2: Smart Grids, Grid Stability, and Offsite and Emergency Power; Advanced and Next Generation Reactors, Fusion Technology; Safety, Security, and Cyber Security; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues
Conference Proceedings	2016 24th International Conference on Nuclear Engineering
Language	English
Publisher Date	2016-06-26
Publisher Place	Charlotte, North Carolina, USA
Access Restriction	Subscribed
Subject Keyword	Temperature distribution Temperature Cooling Computational fluid dynamics Dynamic analysis Neutrons Lithium Ceramics Helium Flow (dynamics) Heat flux Design Fluids Heating China Coolants Thermal conductivity
Content Type	Text
Resource Type	Article

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