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The use of lithium ceramic pebble beds in the design of blankets for fusion reactors makes the mechanical and thermal properties of ceramic pebble beds key issues to be investigated.
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For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronic and thermal design of blanket is under way to find a feasible blanket concept.
Based on the experience of the thermal convection LiPb loops, a multi-functional forced convection loop, DRAGON-IV, was constructed to confirm the different design functions of blankets in China.
The problems range from removing tritium from waste water to the design of thermal blankets necessary to counter intense radiation.
The design of the blankets mainly determines the efficacy of a forced-air warming system.
To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA.
For high performance of the V Li blanket with good economic feature, the design of the blanket should consider to use V 4Cr 4Ti in various thermo-mechanical states in terms of the operation temperature.
This paper presents the conceptual design of the Blanket Remote Handling System (BRHS), which mainly comprises the In-Vessel-Maintenance-System (IVMS), Lifting System and Blanket-Tool-Manipulator System (BTMS).
This work reports on the results of an investigation of the thermal and structural performances of a new design of this blanket, proposed in 2015 by the KIT HCPB Team aimed at establishing a baseline design of the HCPB breeding blanket following the updated EU DEMO plant specifications.
The tritium transport analysis provides valuable reference for the design of HCPB blanket.
A hybrid 3D-1D-3D aproposedis proposed for the conceptual design of a blanket.
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