Neutronics Design And Analysis Of Helium Cooled Solid Breeder Blanket For CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant As its objectives are to demonstrate steady state operation of fusion power and to realize tritium self-sufficiency, the tritium breeding ratio (TBR) is a key design parameter. Because the blankets were designed to withstand the extremely neutron irradiation and to utilize the fusion neutrons for tritium breeding. Most of the issues were related to the neutronics design of the blanket, such as the tritium breeding, nuclear heat production, shielding, irradiation damage of the material and the radioactivity safety, etc.In this paper, a helium cooled solid breeder blanket were proposed for supporting the CFETR design and development. In this concept, radial arranged U-shaped breeding zones are adopted lor simpler structure and to make full use of the breeding zones parallel to first wall (FW) and perpendicular to FW as much as possible for tritium breeding. In the neutronics design work, a detailed 3D neutronics model was built by using of a CAD to Monte Carlo Geometry conversion tool McCad and combined with some manual processing in MCNP. Then based on the neutronics model, the neutron wall loading, TBR distribution, power density of the typical outboard blanket and shielding performance were carried out. According to the assessment of the local TBR variations, the results showed that the global TBR for the preliminary blanket scheme was 1.09, which could meet the tritium self-sufficiency target.Then based on the preliminary blanket scheme, the impact analysis of the design parameters on the tritium breeding performance in CFETR was carried out. These parameters include the breeding zones configuration, the blanket structures such as the FW, cooling plate (CP), stiffening plate (SP), cap and side wall, the numbers and areas of the port and the TBR variation as the Li burnup in the breeding materials.By the adjusting of the pebble bed configuration, the breeding zones was optimized, which showed a well tritium breeding and thermal power distribution performance. Then through the impact analysis of the blanket structures on the tritium breeding performance for the typical outboard blanket, the TBR variation trend as function of these parameters on the global TBR was given. Based on the TBR variation trend and combined with the related thermo-hydraulic and thermo-mechanical analyses results, the blanket was optimized, which the cooling plate was 6mm with 4 tritium breeding zones. By the TBR calculation on the detailed 3D neutronics model for the optimized blanket scheme, the results showed that the assessment of the local TBR variation trend based on the typical outboard blanket was meaningful. And the global TBR variation trend could be assessed according to the related study of the local TBR variations for the typical outboard blanket. Moreover, the impact analyses of areas and numbers of ports on TBR showed that in order to achieve the tritium sufficiency target, the most allowable areas for the mid-plane port and upper port was 11m2 and 16.5m2, respectively. The TBR variation as the burnup of the Li in breeding materials showed that the optimized scheme could satisfy the tritium self-sufficiency when it was operated for 8 years.Finally, the neutronics safety analysis for the optimized typical outboard blanket module was carried out, which include the irradiation damage of the blanket, activation, the shutdown dose distribution when transporting in CASK, and the nuclear waste management for the blanket The irradiation damage showed the optimized blanket could satify the CFETR design limit when it was operated for 8 years. The activation analysis showed that the tritium source dominated the blanket activity in the first 243 years after the opration. And the decay heat in the structures dominated the blanket aflerheat in the early phase after the reactor shutdown. The shutdown dose rate distribution of the typical blanket in the CASK showed that the FW were the dominant component to the total shutdown dose. And at the early time after shutdown, it could be high to 104Sv/h,102Sv/h and 1Sv/h for the cooling time of 1month, lyear and 10 years, respectively. For the conservative remote handing limit (<0.01Sv/h), it required for at least for 35 years to operate the blanket. Thus for reducing the required time after shutdown, the advanced remote handing equipment should be developed. The waste management result showed that most of the blanket components could be recycled after the cooling time of 50 years, but the Li4SO4 should be recycled after the cooling time of 100 years for its decay heat was too high at the cooling time of 50 years. The related analyses results were given as the source and reference to the future safety design work.