High-power gyrotrons for electron cyclotron heating and current drive

In many tokamak and stellarator experiments around the globe that are investigating energy production via controlled thermonuclear fusion, electron cyclotron heating and current drive (ECH&CD) are used for plasma start-up, heating, non-inductive current drive and MHD stability control. ECH will be the first auxiliary heating method used on ITER. Megawatt-class, continuous wave (CW) gyrotrons are employed as high-power millimeter (mm)-wave sources. The present review reports on the worldwide state-of-the-art of high-power gyrotrons. Their successful development during the past years changed ECH from a minor to a major heating method. After a general introduction of the various functions of ECH&CD in fusion physics, especially for ITER, Section 2 will explain the fast-wave gyrotron interaction principle. Section 3 discusses innovations on the components of modern long-pulse fusion gyrotrons (magnetron injection electron gun (MIG), beam tunnel, cavity, quasi-optical output coupler, synthetic diamond output window, single-stage depressed collector) and auxiliary components (superconducting magnets, gyrotron diagnostics, high-power calorimetric dummy loads). Section 4 deals with present megawatt-class gyrotrons for ITER, W7-X, LHD, EAST, KSTAR and JT-60SA, and also includes tubes for moderate pulse length machines as, ASDEX-U, DIII-D, HL-2A, TCV, QUEST and GAMMA-10. In Section 5 the development of future advanced fusion gyrotrons is discussed. These are tubes with higher frequencies for DEMO, multi-frequency (multi-purpose) gyrotrons, stepwise frequency tunable tubes for plasma stabilization, injection-locked and coaxial-cavity multi-megawatt gyrotrons, as well as sub-THz gyrotrons for collective Thomson scattering (CTS). Efficiency enhancement via multi-stage depressed collectors, fast oscillation recovery methods and reliability, availability, maintainability and inspectability (RAMI) will be discussed at the end of this section

Magnetic control of DTT alternative plasma configurations

One of the main challenges concerning next generation tokamaks (such as DEMO) will be the development of a heat and power exhaust system able to withstand the large loads expected in the divertor region. A dedicated Divertor Tokamak Test (DTT) facility has been proposed in the EUROfusion Roadmap, with the aim of testing unconventional solutions, such as advanced magnetic configurations and liquid metal divertors. Magnetic control of alternative plasma configurations, such as the X-Divertor, will play a key role in the solution of the heat exhaust and yet can be a challenging point, due to increased sensitivity introduced by secondary x-points. To overcome the complications introduced by secondary x-points in advanced plasma shapes, magnetic control in DTT is achieved by resolving to the eXtreme Shape Controller, in order to control both the plasma shape and the secondary x-point position

A flexible architecture for plasma magnetic control in tokamak reactors

Plasma magnetic control is one of the core engineering issues to be tackled in a fusion device. Over the last years, model based approaches have been proposed to face this issue, proving their effectiveness and allowing to reduce the time span needed for control testing and validation. The first part of this work is intended to give an overview of the subject, from the historical milestones to the underlying physics; the most common techniques for tokamak plasmas electromagnetic modeling and control are also introduced and discussed. After this introduction, a general architecture for plasma magnetic control in tokamaks is proposed. Finally, the proposed solution is applied to the Experimental Advanced Superconducting Tokamak (EAST) tokamak, where a new plasma magnetic control architecture was developed and implemented during the 2016-2018 experimental campaigns, and to the Japan Torus-60 Super Advanced (JT-60SA) device, which is currently under construction in Japan

A flexible architecture for plasma magnetic control in tokamak reactors

Plasma magnetic control is one of the core engineering issues to be tackled in a fusion device. Over the last years, model based approaches have been proposed to face this issue, proving their effectiveness and allowing to reduce the time span needed for control testing and validation. The first part of this work is intended to give an overview of the subject, from the historical milestones to the underlying physics; the most common techniques for tokamak plasmas electromagnetic modeling and control are also introduced and discussed. After this introduction, a general architecture for plasma magnetic control in tokamaks is proposed. Finally, the proposed solution is applied to the Experimental Advanced Superconducting Tokamak (EAST) tokamak, where a new plasma magnetic control architecture was developed and implemented during the 2016-2018 experimental campaigns, and to the Japan Torus-60 Super Advanced (JT-60SA) device, which is currently under construction in Japan