Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

https://iopscience.iop.org/article/10.1088/0029-5515/47/6/S03/pdf

Chapter 3: MHD stability, operational limits and disruptions

This review discusses the present state-of-the-art of passive high-power microwave components for applications in microwave systems for RF plasma generation and heating, plasma diagnostics, plasma and microwave materials processing, spectroscopy, communication, radar ranging and imaging, and for drivers of next generation high-field-gradient electron-positron linear colliders. The paper reports on high-power components for overmoded high-power transmission systems such as smooth-wall waveguides, HE/sub 11/ hybrid mode waveguides and quasi-optical TEM/sub 00/ beam waveguides. These include various types of mode converters, polarizers, cross-section tapers, bends, mode selective filters, pulse compressors, DC-breaks, directional couplers, beam combiners and dividers, vacuum windows, and instruments for mode analysis. Problems of ohmic attenuation and unwanted conversion to parasitic modes are discussed in detail and rules for alignment requirements are given. In the case of waveguide transmission, this review mainly concentrates on circular waveguide components but also deals with rectangular waveguide.

Passive high-power microwave components

High-power millimeter wave sources are a key enabling technology in fusion energy research. The present state of the art of application of these sources to the areas of heating, current generation, and scattering for diagnostic purposes in fusion plasmas is reviewed. The extrapolation of these applications to future devices and the requirements which they place on sources and transmission lines are also discussed.

Applications of high-power millimeter waves in fusion energy research

Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a state-of-the-art digital real-time control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This unique experimental platform has been used to develop and test new control strategies for three important and reactor-relevant tokamak plasma physics instabilities, including the sawtooth, the edge localized mode (ELM) and the neoclassical tearing mode (NTM). These control strategies offer new possibilities for fusion plasma control and at the same time facilitate studies of the physics of the instabilities with greater precision and detail in a controlled environment. The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface, where q is the safety factor. Exploiting this known physical phenomenon, a sawtooth pacing controller was developed which is able to precisely control the time of appearance of the next sawtooth crash. It was also shown that each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes obtained by the sawtooth pacing controller have been used to study the relationship between sawteeth and NTMs. It is known that post-crash MHD activity can provide the "seed" island for an NTM, which then grows under its neoclassical bootstrap drive. Experiments are shown which demonstrate that the seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have shown effective stabilization of both 3/2 and 2/1 NTMs, and have precisely localized the deposition location that is most effective. Studies of current-profile driven destabilization of tearing modes in TCV plasmas with significant amounts of ECCD show a great sensitivity to details of the current profile, but failed to identify a stationary region in the parameter space in which NTMs are always destabilized. This suggests that transient effects intrinsically play a role. Next to instability control, the simultaneous control of magnetic and kinetic plasma profiles is another key requirement for advanced tokamak operation. While control of kinetic plasma profiles around an operating point can be handled using standard linear control techniques, the strongly nonlinear physics of the coupled profiles complicates the problem significantly. Even more, since internal magnetic quantities are difficult to measure with sufficient spatial and temporal resolution —even after years of diagnostic development— routine control of tokamak plasma profiles remains a daunting and extremely challenging task. In this thesis, a model-based approach is used in which physics understanding of plasma current and energy transport is embedded in the control solution. To this aim, a new lightweight transport code has been derived focusing on simplicity and speed of simulation, which is compatible with the demands for real-time control. This code has been named RAPTOR (RApid Plasma Transport simulatOR). In a first-of-its-kind application, the partial differential equation for current diffusion is solved in real-time during a plasma shot in the TCV control system using RAPTOR. This concept is known in control terms as a state observer, and it is applied experimentally to the tokamak current density profile problem for the first time. The real-time simulation gives a physics-model-based estimate of key plasma quantities, to be controlled or monitored in real-time by different control systems. Any available diagnostics can be naturally included into the real-time simulation providing additional constraints and removing measurement uncertainties. The real-time simulation approach holds the advantage that knowledge of the plasma profiles is no longer restricted to those points in space and time where they are measured by a diagnostic, but that an estimate for any quantity can be computed at any time. This includes estimates of otherwise unmeasurable quantities such as the loop voltage profile or the bootstrap current distribution. In a first closed-loop experiment, an estimate of the internal inductance resulting from the real-time simulation is feedback controlled, independently from the plasma central temperature, by an appropriate mix of co- and counter- ECCD. As a tokamak plasma evolves from one state to another during plasma ramp-up or ramp-down, the profile trajectories must stay within a prescribed operational envelope delimited by physics instabilities and engineering constraints. Determining the appropriate actuator command sequence to navigate this operational space has traditionally been a trial-and-error procedure based on experience of tokamak physics operators. A computational technique is developed based on the RAPTOR code which can calculate these trajectories based on the profile transport physics model, by solving an open-loop optimal control problem. The solution of this problem is greatly aided by the fact that the code returns the plasma state trajectory sensitivities to input trajectory parameters, a functionality which is unique to RAPTOR. This information can also be used to construct linearized models around the optimal trajectory, and to determine the active constraint, which can be used for time-varying closed-loop controller design. This physics-model-based approach has shown excellent results and holds great potential for application in other tokamaks worldwide as well as in future devices.

Real-Time Control of Tokamak Plasmas: from Control of Physics to Physics-Based Control

Clear observations of early triggering of neo-classical tearing modes by sawteeth with long quiescent periods have motivated recent efforts to control, and in particular destabilize, sawteeth. One successful approach explored in TCV utilizes electron cyclotron heating in order to locally increase the current penetration time in the core. The latter is also achieved in various machines by depositing electron cyclotron current drive or ion cyclotron current drive close to the q = 1 rational surface. Crucially, localized current drive also succeeds in destabilizing sawteeth which are otherwise stabilized by a co-existing population of energetic trapped ions in the core. In addition, a recent reversed toroidal field campaign at JET demonstrates that counter-neutral beam injection (NBI) results in shorter sawtooth periods than in the Ohmic regime. The clear dependence of the sawtooth period on the NBI heating power and the direction of injection also manifests itself in terms of the toroidal plasma rotation, which consequently requires consideration in the theoretical interpretation of the experiments. Another feature of NBI, expected to be especially evident in the negative ion based neutral beam injection (NNBI) heating planned for ITER, is the parallel velocity asymmetry of the fast ion population. It is predicted that a finite orbit effect of asymmetrically distributed circulating ions could strongly modify sawtooth stability. Furthermore, NNBI driven current with non-monotonic profile could significantly slow down the evolution of the safety factor in the core, thereby delaying sawteeth.

/pdf/sawtooth-control-in-fusion-plasmas-5e9megfbqt.pdf

Sawtooth control in fusion plasmas

Note: Proc. 19th Symposium on Fusion Technology, Lisbon, Portugal, September 1996, 565 - 568 (1996) Reference CRPP-CONF-1997-029 Record created on 2008-05-13, modified on 2017-05-12

Design and installation of the electron cyclotron wave system for the tcv tokamak

Note: Proceedings of the IAEA Technical Committee Meeting on ECRH Physics and Technology for Fusion Devices and the 11th Joint Workshop on Electron Cyclotron Resonance Heating (EC-11), Oh-Arai, Japan, October 1999, JAERI-memo 12-041, 275 - 281 (March 2000) Reference CRPP-CONF-2000-031 Record created on 2008-05-13, modified on 2017-12-10

Poloidally asymmetric plasma response with ECH deposition near q=1 in TCV

Note: Paper Th/P3-10 Reference CRPP-CONF-2006-115 URL: http://www-pub.iaea.org/MTCD/Meetings/Announcements.asp?ConfID=149 Record created on 2008-05-13, modified on 2017-05-12

Partial Stabilization and Control of Neoclassical Tearing Modes in Burning Plasmas

Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne,Association EURATOM-Confederation Suisse, CH-1015 Lausanne EPFL, Switzerland Introduction: Plasma experiments using absorption of radiation in the range of the electroncyclotron frequency are often grouped under two separate headings: electron cyclotron heating(ECH) or electron c yclotron current dri ve (ECCD). The division is based, roughly , on the toroi-dal angle-of-incidence between the beam and the toroidal magnetic ﬁeld direction. In practice,ECH experiments are those in which there is no projection of the k-vector on the toroidal ﬁeld, B

Poloidally asymmetric plasma response during ECH experiments in TCV

The Advanced Tokamak scenario, one of the modes of operation being considered for ITER, relies on the attainment of a high bootstrap current fraction. This scenario is typically characterized by an internal transport barrier (ITB). An internal feedback loop then governs the current profile, which strongly affects the confinement and thus the properties and location of the high-gradient region, where in turn the bootstrap current component is localized. The bootstrap current fraction can reach 100% only if the bootstrap current profile can be exactly and stably aligned with the high-gradient region it engenders. Recent work on the TCV tokamak has shown that such an alignment is indeed possible. We have produced discharges in which the current is entirely self-generated by the plasma in conditions of intense electron cyclotron heating (ECH, up to 2.7 MW), by employing two different methods. In one scenario, high-power, second-harmonic ECH waves are launched with no current drive component, during the initial plasma current ramp-up. A strong ITB is generated by the transient negative central magnetic shear that develops in the current penetration phase. The Ohmic flux swing is zeroed immediately after breakdown, cutting off the external plasma current source. The plasma can then evolve spontaneously towards a stationary and quiescent state, characterized by a narrow ITB with a confinement enhancement of 2.5-3 over Lmode. The discharge remains stationary over the time scale of a TCV pulse (1-2 s), which is significantly longer than a typical resistive current redistribution time ( 150-300 ms) and orders of magnitude longer than the confinement time ( 3-6 ms). Following a different approach, we have also succeeded in achieving a 100% bootstrap fraction by annulling the total EC-driven current in pre-existing stationary conditions. Standard stationary noninductive eITBs are first generated by off-axis co-ECCD; counter-ECCD is then added gradually until the total driven current density is nominally zero everywhere. In some cases a quasi-stationary state is indeed established. In this case the barrier remains broad with a standard confinement enhancement factor of the order of 4-5.

/pdf/full-bootstrap-discharge-sustainment-in-steady-state-in-the-4jrjhmsils.pdf

Full Bootstrap Discharge Sustainment in Steady State in the TCV Tokamak

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