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• The preliminary design of helium-cooled mixed bed breeder (HCMB) blanket for CFETR is performed. • The neutronics and thermal-hydraulics analyses of CFETR are performed. • The global TBR of HCMB is larger than that of HCCB. Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device designed with different power levels to validate fusion engineering technologies in China. Helium-cooled ceramic breeder (HCCB) blanket and Water-cooled ceramic breeder (WCCB) blanket are the main candidates. Compared with WCCB, HCCB generally has a higher tritium breeding ratio (TBR) because helium has smaller neutron absorption cross sections than water. However, the TBR of the HCCB decreases with the increase of the fusion power. In this paper, the preliminary neutronics and thermal-hydraulics design of the Helium-Cooled Mixed bed Breeder (HCMB) blanket is proposed. The blanket structure with a fusion power of 1.5 GW is designed based on the current HCCB design of the CFETR. The HCMB adopts helium as the coolant and the mixed Li 4 SiO 4 /Be 12 Ti as the tritium breeder. The neutronics and thermal-hydraulics performances of HCMB blanket have been analyzed in the paper. The results show that the HCMB blanket design has a TBR up to 1.25 with the thermal-hydraulics requirements satisfied. [ABSTRACT FROM AUTHOR]

To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and is being designed mainly to demonstrate 50–200 MW fusion power, 30–50% duty time factor, tritium self-sustained. Because of the high demand of tritium production and the realistic engineering consideration, the design of tritium breeding blanket for CFETR is a challenging work and getting special attention. As a blanket candidate, a helium cooled solid breeder blanket has been designed with the emphasis on conservative design and realistic blanket technology. This paper introduces the basic blanket scheme, including the material selection, structural design, cooling scheme and purge gas flow path. In addition, some results of neutronics, thermal-hydraulic and stress analysis are presented. [ABSTRACT FROM AUTHOR]

• A complete nuclear analysis of the CFETR DFLL has been carried out at INEST with SuperMC Monte Carlo code and JEFF3.2. • The tritium production was improved to meet the self-sufficiency of CFETR. • The reference design met all shielding criteria (fast neutron flux, nuclear heating, DPA). China Fusion Engineering Test Reactor (CFETR), as one of the option fusion reactors in China, has been proposed to fill the gap between ITER and fusion DEMO plant. In order to evaluate the neutronics performance in the framework of National Magnetic Confinement Fusion Science Program of China, the neutronics analyses of the detailed model of CFETR with China dual-functional lithium lead (DFLL) test blanket module (TBM) were carried out with the Super Multi-functional Calculation Program for Nuclear Design Safety Evaluation (SuperMC) and JEFF-3.2 nuclear data library. To meet the self-sufficiency target of CFETR, the tritium breeding ratio has been improved to 1.2. Meanwhile, the nuclear heating result showed a potential for energy multiplication factor improvement. From this previous study, an optimized model was proposed to improve the shielding performance and the results showed that the DFLL-TBM with the optimized model could meet the tritium self-sufficiency target and radiation design requirements. [ABSTRACT FROM AUTHOR]