The control and data acquisition system of the DTT experiment

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
G. Manduchi , F. Zanon , L. Boncagni , P. Mosetti , G. Martini , G. Paccagnella , C. Centioli , R. Ambrosino , F. Sartori
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引用次数: 0

Abstract

DTT, Divertor Tokamak Test facility, is currently under construction at the Frascati ENEA Research Center. Its main aim is to explore alternative solutions for the extraction of the heat generated by the fusion process. Its Control and Data Acquisition System (CODAS) will (1) orchestrate and synchronize all the DTT systems during Plasma operation and maintenance; (2) acquire data from the experiment diagnostics and plant systems and store it in an experimental database to be used for on-line and off-line analysis; (3) provide real-time Plasma control. The expected duration of the plasma discharge in DTT is in the order of some tens of seconds and therefore DTT can be considered a long-lasting experiment, involving data streaming technologies for data communication and storage. The main DTT CODAS design is based on three principles: (1) Taking inspiration from other similar experiments currently under development, namely ITER CODAC, (2) relying on proven solutions already adopted in running experiments with similar constraints and (3) taking advantage from practices widely adopted in fusion and, more in general, in industry. Despite using components already adopted in other fusion experiments, DTT CODAS is the first system that seamlessly integrates all of them.
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来源期刊
Fusion Engineering and Design
Fusion Engineering and Design 工程技术-核科学技术
CiteScore
3.50
自引率
23.50%
发文量
275
审稿时长
3.8 months
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.
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