Monika Lewandowska , Aleksandra Dembkowska , Rafał Ortwein , Arend Nijhuis , Giulio Anniballi , Lorenzo Giannini , Danko C. van der Laan , Jeremy Weiss , Gianluca De Marzi , Davide Uglietti
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引用次数: 0
Abstract
Conductors using High Temperature Superconductor (HTS) tapes are considered as a very promising solution for future high-field fusion magnets. Various HTS cable concepts, such as e.g. twisted stack cable, cross-conductor (CroCo), Roebel assembled coated conductor (RACC), conductor on round core (CORCⓇ), HTS cable-in-conduit conductor (CICC), aligned stacks transposed in Roebel arrangement (ASTRA) have been proposed. Some of them are already considered for potential use in some components of the EU-DEMO magnet system. Recently a design of a conductor based on the CORCⓇ concept for the innermost layer of the hybrid Central Solenoid (CS) coil of EU-DEMO was proposed. In the present work we simulated, using the THEA code by Cryosoft, normal operation of this conductor during the current cycle. Two variants of the CS1 geometry were considered: a single CS1 module, and a CS1 module split into two sub-modules (CS1L and CS1U) located one above the other. Taking into account heat loads due to the hysteresis and coupling losses, resulting from time evolution of the magnetic field profile along the conductor, we estimated the minimum temperature margin in the conductor, to verify if it fulfills the performance criterion: min(ΔTmarg) > 1.5 K. The results obtained with the 1D THEA model were complemented by the results of simulations with the ANSYS Mechanical APDL model (3D solid + 1D fluid) aimed at the assessment of radial temperature gradients in the conductor cross-section.
期刊介绍:
The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.