A.S. Welander

Advanced techniques for neoclassical tearing mode control in DIII-D

Two techniques were developed at DIII-D [J. L. Luxon, Nucl. Fusion 42, 64 (2002)] to tackle ITER-specific aspects of neoclassical tearing mode (NTM) control, namely, (1) the relatively small size of the rotating islands, smaller than the electron cyclotron current drive (ECCD) deposition region, and (2) the increased tendency of the islands, compared to present devices, to lock to the wall or to the residual error field, in a position not necessarily accessible to ECCD. Modulated ECCD is known to suppress small islands more efficiently, when “broad,” than continuous ECCD. At DIII-D, a NTM of poloidal/toroidal mode numbers m/n=3/2 was completely stabilized by a new technique where oblique electron cyclotron emission acted at the same time as an indicator of good alignment between ECCD and the island, and as a waveform generator, for modulation in synch and in phase with the island O-point. In another experiment, after locking in an unfavorable position, a 2/1 island was steered by externally generated magn...

Experimental vertical stability studies for ITER performance and design guidance

United States Department of Energy (DE-FC02-04ER54698, DEAC52- 07NA27344, and DE-FG02-04ER54235)

Active control for stabilization of neoclassical tearing modes

This work describes active control algorithms used by DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to stabilize and maintain suppression of 3/2 or 2/1 neoclassical tearing modes (NTMs) by application of electron cyclotron current drive (ECCD) at the rational q surface. The DIII-D NTM control system can determine the correct q-surface/ECCD alignment and stabilize existing modes within 100–500ms of activation, or prevent mode growth with preemptive application of ECCD, in both cases enabling stable operation at normalized beta values above 3.5. Because NTMs can limit performance or cause plasma-terminating disruptions in tokamaks, their stabilization is essential to the high performance operation of ITER [R. Aymar et al., ITER Joint Central Team, ITER Home Teams, Nucl. Fusion 41, 1301 (2001)]. The DIII-D NTM control system has demonstrated many elements of an eventual ITER solution, including general algorithms for robust detection of q-surface/ECCD alignment and for real-time maintenance of alignment ...

Development of ITER 15 MA ELMy H-mode inductive scenario

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation.

Comprehensive control of resistive wall modes in DIII-D advanced tokamak plasmas

The resistive wall mode (RWM) and neoclassical tearing mode (NTM) have been simultaneously suppressed in the DIII-D for durations of over 2 s at beta values 20% above the no-wall limit with modest electron cyclotron current drive and very low plasma rotation. The achieved plasma rotation was significantly lower than reported previously. However, in this regime where stable operation is obtained, it is not unconditionally guaranteed. Various MHD activities, such as edge localized modes (ELMs) and fishbones, begin to couple to the RWM branch near the no-wall limit; feedback has been useful in improving the discharge stability to such perturbations. Simultaneous operation of slow dynamic error field correction and fast feedback suppressed the pile-up of ELM-induced RWM at a series of ELM events. This result implies that successful feedback operation requires not only direct feedback against unstable RWM but also careful control of MHD-induced RWM aftermath, which is the dynamical response to a small-uncorrected error field near the no-wall beta limit. These findings are extremely useful in defining the challenge of control of the RWM and NTM in the unexplored physics territory of burning plasmas in ITER.

Development of ITER-relevant plasma control solutions at DIII-D

The requirements of the DIII-D physics program have led to the development of many operational control results with direct relevance to ITER. These include new algorithms for robust and sustained stabilization of neoclassical tearing modes with electron cyclotron current drive, model-based controllers for stabilization of the resistive wall mode in the presence of ELMs, coupled linear–nonlinear algorithms to provide good dynamic axisymmetric control while avoiding coil current limits, and adaptation of the DIII-D plasma control system (PCS) to operate next-generation superconducting tokamaks. Development of integrated plasma control (IPC), a systematic approach to modelbased design and controller verification, has enabled successful experimental application of high reliability control algorithms requiring a minimum of machine operations time for testing and tuning. The DIII-D PCS hardware and software and its versions adapted for other devices can be connected to IPC simulations to confirm control function prior to experimental use. This capability has been important in control system implementation for tokamaks under construction and is expected to be critical for ITER.

Stabilization and prevention of the 2/1 neoclassical tearing mode for improved performance in DIII-D

The m = 2/n = 1 neoclassical tearing mode (NTM) has been observed to strongly degrade confinement and frequently lead to a disruption in high β discharges in DIII-D if allowed to grow to a large size. Stabilization of grown NTMs by the application of a highly localized electron cyclotron current drive (ECCD) at the island location has led to operation at an increased plasma pressure, up to the no-wall kink limit. After the NTM is stabilized by the ECCD, the correct location for the current drive is maintained using information from real-time equilibrium reconstructions which include measurements from the motional Stark effect diagnostic. This same process is used alternatively to prevent the mode from ever growing, leading to performance at the pressure limit in high performance hybrid discharges with β above 4%. Modelling using the modified Rutherford equation shows that the required power is in close agreement with the experimental threshold for the prevention of the 2/1 NTM.

State-of-the-art neoclassical tearing mode control in DIII-D using real-time steerable electron cyclotron current drive launchers

Real-time steerable electron cyclotron current drive (ECCD) has been demonstrated to reduce the power requirements and time needed to remove 3/2 and 2/1 neoclassical tearing modes (NTMs) in the DIII-D tokamak. In a world first demonstration of the techniques required in ITER, the island formation onset is detected automatically, gyrotrons are turned on and the real-time steerable ECCD launcher mirrors are moved promptly to drive current at the location of the islands. This shrinks and suppresses the modes well before saturation using real-time motional Stark effect constrained equilibria reconstruction with advanced feedback and search algorithms to target the deposition. In ITER, this method will reduce the ECCD energy requirement and so raise Q by keeping the EC system off when the NTM is not present. Further, in the experiments with accurate tracking of pre-emptive ECCD to resonant surfaces, both 3/2 and 2/1 modes are prevented from appearing with much lower ECCD peak power than required for removal of a saturated mode.

Higher stable beta by use of pre-emptive electron cyclotron current drive on DIII-D

Electron cyclotron current drive (ECCD) is used in conjunction with accurate real-time equilibrium reconstructions to operate a tokamak plasma at high beta without the destabilization of a performance-degrading neoclassical tearing mode that otherwise is metastable and therefore appears consistently. This is the first experiment in which the alignment of the ECCD on the rational surface being stabilized is maintained in the absence of the mode. Driving current at the rational surface eliminates the metastable condition, thereby making the mode stable.

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.