L. Zabeo

Novel aspects of plasma control in ITER

ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

Present status of the ITER real-time Plasma Control System development

ITER will be the worlds largest magnetic confinement tokamak fusion device and is currently under construction in southern France. The ITER Plasma Control System (PCS) is a fundamental component of the ITER Control, Data Access and Communication system (CODAC). It will control the evolution of all plasma parameters that are necessary to operate ITER throughout all phases of the discharge. The design and implementation of the PCS poses a number of unique challenges. The timescales of phenomena to be controlled spans three orders of magnitude, ranging from a few milliseconds to seconds. Novel control schemes, which have not been implemented at present-day machines need to be developed, and control schemes that are only done as demonstration experiments today will have to become routine. In addition, advances in computing technology and available physics models make the implementation of real-time or faster-than-real-time calculations to forecast and subsequently to avoid disruptions or undesired plasma regimes feasible. A further novel feature is a sophisticated event handling system, which provides a means to deal with plasma related events (such as MIlD instabilities or LH transitions) or component failure. Finally, the schedule for design and implementation poses another unique challenge. The beginning of ITER operation will be in late 2020, but the conceptual design activity has already commenced as required by the on-going development of diagnostics and actuators in the domestic agencies and the need for integration and testing. In this paper, an overview about the functional requirements for the plasma control system will be given. The main focus will be on the requirements and possible options for a real-time framework for ITER and its interfaces to other ITER CODAC systems (networks, other applications, etc.). The limited amount of commissioning time foreseen for plasma control will make extensive testing and validation necessary. This should be done in an environment that is as close to the PCS version running the machine as possible. Furthermore, the integration with an Integrated Modeling Framework will lead to a versatile tool that can also be employed for pulse validation, control system development and testing as well as the development and validation of physics models. An overview of the requirements and possible structure of such an environment will also be presented.

Jet operations and plasma control: A plasma control system that is safe and flexible in a manageable way.

This paper presents the main lessons that can be learned from two decades of Plasma Control development experience. It will present the general architectural choices and will provide examples from key systems, and, more importantly, will highlight the plasma operation requirements that have driven the developments. The aim is to present a meaningful set of functional and non- functional requirements that were derived from the JET experience and to discuss their potential applicability to future experimental devices, ITER in particular.

Plasma vertical stabilisation in ITER

This paper describes the progress in analysis of the ITER plasma vertical stabilisation (VS) system since its design review in 2007–2008. Two indices characterising plasma VS were studied. These are (1) the maximum value of plasma vertical displacement due to free drift that can be stopped by the VS system and (2) the maximum root mean square value of low frequency noise in the dZ/dt measurement signal used in the VS feedback loop. The first VS index was calculated using the PET code for 15 MA plasmas with the nominal position and shape. The second VS index was studied with the DINA code in the most demanding simulations for plasma magnetic control of 15 MA scenarios with the fastest plasma current ramp-up and early X-point formation, the fastest plasma current ramp-down in a divertor configuration, and an H to L mode transition at the current flattop. The studies performed demonstrate that the VS in-vessel coils, adopted recently in the baseline design, significantly increase the range of plasma controllability in comparison with the stabilising systems VS1 and VS2, providing operating margins sufficient to achieve ITERs goals specified in the project requirements. Additionally two sets of the DINA code simulations were performed with the goal of assessment of the capability of the PF system with the VS in-vessel coils: (i) to control the position of runaway electrons generated during disruptions in 15 MA scenarios and (ii) to trigger ELMs in H-mode plasmas of 7.5 MA/2.65 T scenarios planned for the early phase of ITER operation. It was also shown that ferromagnetic structures of the vacuum vessel (ferromagnetic inserts) and test blanket modules insignificantly affect the plasma VS.

Control, detection and mitigation of disruptions on ITER

The fast loss of thermal and magnetic energy during an unmitigated disruption in ITER can lead to heat fluxes exceeding melt thresholds for plasma facing components, can generate high electromagnetic loads in some cases close to the design limits, and can potentially cause the generation of high energy runaway electrons. Therefore, the operation strategy in ITER will have a strong focus on limiting the number of disruptions and especially of those that are unmitigated. This will be achieved by a multi-layer strategy including disruption prevention and avoidance, disruption prediction and disruption mitigation. The plasma control system (PCS) will be in charge to react to any deviation from the predicted behavior by applying appropriate correction measures for which an extensive suite of actuators is available including those for active disruption prevention like electron cyclotron heating. It is also the PCS that optimizes triggering of the disruption mitigation system (DMS), should a disruption be imminent despite all control efforts. This paper will give an overview on all aspects of disruption control with a more detailed discussion of the requirements for the DMS.

Real-time identification of the current density profile in the JET Tokamak: method and validation

The real-time reconstruction of the plasma magnetic equilibrium in a Tokamak is a key point to access high performance regimes. Indeed, the shape of the plasma current density profile is a direct output of the reconstruction and has a leading effect for reaching a steady-state high performance regime of operation. In this paper we present the methodology followed to identify numerically the plasma current density in a Tokamak and its equilibrium. In order to meet the real-time requirements a C++ software has been developed using the combination of a finite element method, a nonlinear fixed point algorithm associated to a least square optimization procedure. The experimental measurements that enable the identification are the magnetics on the vacuum vessel, the interferometric and polarimetric measurements on several chords and the motional Stark effect. Details are given about the validation of the reconstruction on the JET tokamak, either by comparison with ‘off-line’ equilibrium codes or real time software computing global quantities.