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There is an increasing need for integrating individual plasma-control algorithms with the ultimate goal of simultaneously regulating more than one plasma property. Some of these integrated-control solutions should have the capability of arbitrating the authority of the individual plasma-control algorithms over the available actuators within the tokamak. Such decision-making process must run in real time since its outcome depends on the plasma state. Therefore, control architectures including supervisory and/or exception-handling algorithms will play an essential role in future fusion reactors like ITER. However, most plasma-control experiments in present devices have focused so far on demonstrating control solutions for isolated objectives. In this work, initial experimental results are reported for simultaneous current-profile control, normalized-beta control, and Neoclassical Tearing Mode (NTM) suppression in DIII-D. Neutral beam injection (NBI), electron-cyclotron (EC) heating & current drive (H&CD), and plasma current modulation are the actuation methods. The NBI power and plasma current are always modulated by the Profile Control category within the DIII-D Plasma Control System (PCS) in order to control both the current profile and the normalized beta. EC H&CD is utilized by either the Profile Control or the Gyrotron categories within the DIII-D PCS as dictated by the Off-Normal and Fault Response (ONFR) system, which monitors the occurrence of an NTM and regulates the authority over the gyrotrons. The total EC power and poloidal mirror angles are the gyrotron-related actuation variables. When no NTM suppression is required, the gyrotrons are used by the Profile Control category, but when NTM suppression is required, the ONFR transfers the authority over the gyrotrons to the NTM stabilization algorithm located in the Gyrotron category. Initial experimental results show that simultaneous control of different aspects of the plasma dynamics may improve the overall control and plasma performances. Also, the potential of the ONFR system to successfully integrate competing control algorithms is demonstrated.

Fast vertical control of shaped plasmas is essential for the successful realization of plasma operation near the limit of vertical control to achieve maximum confinement and stable, disruption free operation. In KSTAR, the normal conducting in-vessel vertical control coil (IVC coil) is employed to respond to vertical transients much faster than the superconducting coils are capable. The power supply for the IVC coil is capable of responding in a time commensurate with the expected vertical growth rates on KSTAR. The diagnostics used for the fast vertical control since the first operation with shaped plasmas in KSTAR, however, have a low signal-to-noise ratio which limits the gains that can be used successfully in the proportional and derivative control loops and thus the speed of the control loop. The successful use of relative flux for the Z-position estimate and of the loop voltage difference from a pair of up-down symmetric loops to provide sensitive and less noisy vertical estimate will be discussed. In addition, the control loop for the IVC coil is decoupled from the slow motion controlled by the superconducting coils using a high-pass filter in the control software. Finally the noticeable improvement in plasma control for plasmas near the vertical stability limit will be discussed.

Divertor detachment offers a promising solution to the challenge of plasma-wall interactions for steady-state operation of fusion reactors. Here, we demonstrate the excellent compatibility of actively controlled full divertor detachment with a high-performance (βN ~ 3, H98 ~ 1.5) core plasma, using high-βp (poloidal beta, βp > 2) scenario characterized by a sustained core internal transport barrier (ITB) and a modest edge transport barrier (ETB) in DIII-D tokamak. The high-βp high-confinement scenario facilitates divertor detachment which, in turn, promotes the development of an even stronger ITB at large radius with a weaker ETB. This self-organized synergy between ITB and ETB, leads to a net gain in energy confinement, in contrast to the net confinement loss caused by divertor detachment in standard H-modes. These results show the potential of integrating excellent core plasma performance with an efficient divertor solution, an essential step towards steady-state operation of reactor-grade plasmas., Plasma fusion devices like tokamaks are important for energy generation but there are many challenges for their steady state operation. Here, the authors show that full divertor detachment is compatible with high-confinement high-poloidal-beta core plasmas and this prevents the damage to the divertor target plates and the first wall.

Future tokamaks will require robust technologies for the mitigation of heat exhaust onto the plasma-facing components. As a first step towards this development, a system has been developed at DIII-D that estimates and controls in real-time the heat flux to the PFCs. Real-time estimation of the peak power flux from this model-based approach is validated with off-line infra-red measurements for various DIII-D plasma discharges. A nonlinear free-boundary simulation code (GSevolve) is used for simulating the closed loop response and for the off-line determination of the control parameters. The implementation and first experimental results of the application during the DIII-D plasma campaign are reported.

To investigate the optimal scenario of impurity seeding to obtain divertor plasma detachment for target protection, experiments with Ar&D2 seeding from two different poloidal locations, the upper outer (UO) divertor target and lower outer (LO) target, were carried out on EAST. Partial energy detachment (the electron temperature near the strike point T e,spt ⩽ 10 eV) were obtained with Ar&D2 mixture puffing from the UO target and, for the first time, from the LO target into H-mode plasmas in the upper single null (USN) configuration. The peak heat flux q t on the UO target was significantly reduced (by ∼80%). The rollover of ion flux density j s did not appear probably due to insufficient momentum loss, which is independent of the puff locations. The poloidal asymmetries of particle and heat fluxes on the targets have also been investigated. The UO-dominant asymmetry of particle flux was reversed, while the UO-dominant asymmetry of heat flux was mitigated but not reversed. The plasma confinement dropped by only 14% and 8.2% in the LO-puff case and UO-puff case, respectively, during detachment. The high level of C may contributed greatly to the higher radiation in the bulk plasma region and the greater decline in W MHD in the LO-puff case.

Integration of divertor detachment with a high-performance (βN ∼ 3, βp > 2, H98 ∼ 1.5) core plasma has been demonstrated in DIII-D high-βp (poloidal beta) plasmas associated with a sustained core internal transport barrier (ITB) and an H-mode edge transport barrier (ETB). Such good core-edge integration has been achieved for both neon and nitrogen seeding, for both favorable and unfavorable B-field directions, independently from the impurity puffing locations, though these variations play important roles on divertor characteristics. Compared to the standard H-mode plasmas, the high-βp plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower $n_{e}$ , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative $D_{2}$ -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative $D_{2}$ -seeded SF divertor at $P_{\text {SOL}}=3-4$ MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected $P_{\text {SOL}} \simeq 9$ MW case. The radiative SF divertor with carbon impurity provides a wider $n_{e}$ operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar $q_{\text {peak}}$ reduction factors (see standard divertor).

The plasma control system (PCS) has been one of essential systems in annual KSTAR plasma campaigns: starting from a single-process version in 2008, extensive upgrades are done through the previous 7 years in order to achieve major goals of KSTAR performance enhancement. Major implementations are explained in this paper. In consequences of successive upgrades, the present KSTAR PCS is able to achieve ∼48 s of 500 kA plasma pulses with full real-time shaping controls and real-time NB power controls. It has become a huge system capable of dealing with 8 separate categories of algorithms, 26 actuators directly controllable during the shot, and real-time data communication units consisting of +180 analog channels and +600 digital input/outputs through the reflective memory (RFM) network. The next upgrade of the KSTAR PCS is planned in 2015 before the campaign. An overview of the upgrade layout will be given for this paper. The real-time system box is planned to use the CERN MRG-Realtime OS, an ITER-compatible standard operating system. New hardware is developed for faster real-time streaming system for future installations of actuators/diagnostics.

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated 'SAS' effects and cross-field drifts, favouring operation with the ion B × ∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D's open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with T e ≲ 10 eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B × ∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors.

Plasma discharges with a negative triangularity (δ=-0.4) shape have been created in the DIII-D tokamak with a significant normalized beta (β_{N}=2.7) and confinement characteristic of the high confinement mode (H_{98y2}=1.2) despite the absence of an edge pressure pedestal and no edge localized modes (ELMs). These inner-wall-limited plasmas have a similar global performance as a positive triangularity (δ=+0.4) ELMing H-mode discharge with the same plasma current, elongation and cross sectional area. For cases both of dominant electron cyclotron heating with T_{e}/T_{i}>1 and dominant neutral beam injection heating with T_{e}/T_{i}=1, turbulent fluctuations over radii 0.5

Feedback control of radiated power from the lower divertor P rad, div, L has been implemented in the DIII-D Plasma Control System (PCS). A realtime sensor for P rad, div, L has been constructed from 12 foil bolometer channels which agrees with standard post-shot analysis to within 20%. Results with the 12-channel sensor are compared to initial proof-of-concept tests with a single channel as a proxy for P rad, div, L , showing that the upgraded sensor is necessary to overcome limitations of the proxy channel strategy in DIII-D. Using N 2 seeding under feedback control, P rad, div, L has been increased by up to 150% above unseeded levels, and a radiated power fraction f rad of 80% has been demonstrated, although feedback controlled gas flow is steadier at f rad = 55%. Spatial coverage is broad enough to enable P rad control during the strike point sweeps which are commonly used to generate pseudo-2D divertor Thomson measurements in DIII-D divertor experiments. Use of this control reveals challenges that may affect next step devices, which will require actively controlled extrinsic impurity seeding in order to manage heat loads. When operating at high f rad , changes in pedestal T e (caused by ELMs in these experiments but could come from other disturbances) resulted in large perturbations to P rad which were destabilizing to the feedback controller.

The exhaust of excessively high heat and particle fluxes on the divertor target is crucial for EAST long-pulse operation. In the recent EAST experiments, stable partial energy detachment around the upper outer strike point with H 98,y2 ∼ 1 was achieved with either Ne or Ar seeding from the upper outer divetor target in the upper single null configuration with ITER-like tungsten divertor. With either Ar or Ne seeding, the electron temperature around the upper outer strike point (T et,UOSP) was maintained at around 5 eV, the peak temperature of divertor target surface around the upper outer strike point (T div,UO) decreased significantly, and material sputtering was well suppressed. It was observed that there was less Ar seeding needed for partial energy detachment onset than Ne seeding, which shows that Ar is more efficient in the cooling of T et on the upper outer divertor than Ne. However, there was no detachment on the upper inner divertor with T et around strike point (T et,UISP) remaining >10 eV with either Ar or Ne seeding from the upper outer divertor. Accompanied with the disappearance of double peak phenomenon of ion flux density on the upper inner divertor target (j s,UI), the peak T div,UI around the strike point increased to around 300 °C. Although the heat flux on the upper inner divertor target (q t,UI) is still in the acceptable level, either Ar or Ne seeding only from the upper outer divertor target is not enough to protect the upper inner divertor target from sputtering under current EAST conditions. On the other hand, Ar seeding always causes confinement degradation in the partial energy detachment state. It was observed that there is a slight confinement improvement (∼10%) with Ne seeding, which may be due to density peaking, dilution effects and stabilization of the ion temperature gradient mode.

Experimental and modeling investigations of high βp scenarios on DIII-D and EAST tokamaks show advantages in high energy confinement, avoidance of n = 1 MHD, and core-edge integration with reduced heat flux, making this scenario an attractive option for China Fusion Engineering Test Reactor steady-state operation. Experiments show that plasmas with high confinement and high density can be achieved with neutral beam injection on DIII-D (βp ∼ 2.2, βN ∼ 3.5, fBS ∼ 50%, fGw ∼ 1.0, and H98y2 ∼ 1.5) and pure RF power on EAST (βP ∼ 2.0, βN ∼ 1.6, fBS ∼ 50%, fGw ∼ 0.8, and H98y2 > 1.3). By tailoring the current density profile, a q-profile with local (off-axis) negative shear is achieved, which yields improved confinement and MHD stability. Transport analysis and simulation suggest that the combination of a high density gradient and high Shafranov shift allows turbulence stabilization and higher confinement. Using on-axis Electron Cyclotron Heating injection, tungsten accumulation is avoided on EAST, and this is reproduced in modeling. Reduced heat flux (by > 40%) and maintenance of high core confinement is achieved with active feedback control of the radiated divertor, an important result for long pulse operation in tokamaks. The improved physics understanding and validated modeling tools are used to design a 1 GW steady-state scenario for CFETR.

A new controller has been developed with help of the flexible divertor poloidal-field coil set of the DIII-D tokamak, to aid in the precise control of the flux expansion in the scrape-off layer. The single-input multiple-output architecture ensures flexibility through a complementary set of orthogonal actuator direction to guarantee minimum effect on existing controlled variables, e.g. radial and vertical position control of the X-point. A non-linear free-boundary simulation code (GSevolve) is used for simulating the closed-loop response and for verifying the implementation of the control algorithm on the DIII-D plasma control system. First results of the experimental commissioning of the new controller during 2020 DIII-D campaign are also presented.

Improvements in strike point control enabled tests with the new small angle slot (SAS) divertor installed at DIII-D with precision as tight as ± 1 mm, barring brief perturbations from ELMs. Aside from being narrower than previous baffle structures in DIII-D, simulations indicate that performance of the SAS divertor should be very sensitive to strike point position, making the improved strike point control essential for experiments with the SAS divertor. Improved control entails regulating more divertor geometry target locations than before and setting higher gain on the strike point target. This is achieved by tasking more poloidal field (PF) shaping coils with divertor topology control, fine tuning of the coil configuration and algorithm gains, and adding regulation of the power supply common rail. The result is that although more variables are controlled, precision is better: high frequency jitter in the strike point position is lower on average compared to related discharges with standard divertor shape control.

An overview of advances and progress on the KSTAR plasma control improvement is given. The enhancement of the KSTAR plasma control system (PCS) continues in order to enable implementations of more sophisticated control algorithms and capabilities of integrated controls on magnets, gas, heating and off-normal event handling. Features and functionalities on the generic magnetic controls, kinetic controls and new fault scheme design are added for the operation space expansion, mainly regarding recent achievements of high plasma current up to 1.1 MA for 10 s, high-performance long pulse up to 90 s, and discharges relevant to ITER research. Technical challenges and prospective upgrades in the near future are described.

EAST achieved 101.2 seconds H-mode discharge with total power injection up to 3MW in 2017. Consequently to the enhanced auxiliary heating power, the active divertor heat flux control is urgently needed in EAST. In 2018, the active divertor detachment feedback control was implemented successfully in EAST plasma control system (PCS) with divertor Langmuir probe measurements for evaluating the divertor particle flux. The control algorithm controlled the degree of detachment effectively during the 2018 EAST experiments through either D2 fueling in the low field side using SMBI or impurity seeding from the divertor volume in H-mode plasmas, without significant degradation of the core-plasma performance, i.e., exhibiting good core-edge-divertor integration. The algorithm implementation in PCS and experimental results will be reported in this paper. Further efforts will be made to extend active feedback control towards long-pulse H-mode detachment maintenance in the near future, which will benefit the high power, H-mode operation for ITER.

An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D 2 D 2 seeding. The SF configuration was maintained for many energy confinement times (2–3 s) in H-mode discharges ( I p =1.2 I p = 1.2 MA, P NBI =4–5 P NBI = 4 – 5 MW, and B×∇B B × ∇ B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary B p B p null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

Sustaining the plasma in a stable and a high performance condition is one of the important control issues for future steady state tokamaks. In the 2014 KSTAR campaign, we have developed a real-time poloidal beta (βp) control technique and carried out preliminary experiments to identify its feasibility. In the control system, the βp is calculated in real time using the measured diamagnetic loop signal, and compared with the target value leading to the change of the neutral beam (NB) heating power using a feedback PID control algorithm. To match the requested power of NB which is operated with constant voltage, the on-time periods of the intervals were adjusted as the ratio of the required power to the maximum achievable one. This paper will present the overall procedures of the βp control, the βp estimation process and NB operation scheme implemented in the plasma control system (PCS), and the analysis on the preliminary experimental results.

An accelerated fusion energy development program, a fast-track approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation,and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 κ = 2.3, BT= 5.4 T, IP = 6.6 MA, βN = 2.75, Pfus = 127 MW. The modest bootstrap fraction of fBS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Lastly,more » proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ~ 10 in ITER.« less

Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (Ip) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16 s) with the number of required commissioning shots largely reduced.

The DIII-D plasma control system (PCS), initially deployed in the early 1990s, now controls nearly all aspects of the tokamak and plasma environment. Versions of this PCS, supported by General Atomics, are presently used to control several tokamaks around the world, including the superconducting tokamaks Experimental Advanced Superconducting Tokamak and Korean Superconducting Tokamak Advanced Research. The experimental challenges posed by the advanced tokamak mission of DIII-D and the variety of devices supported by the PCS have driven the development of a rich array of control algorithms, along with a powerful set of tools for algorithm design and testing. Broadly speaking, the PCS mission is to utilize all available sensors, measurements, and actuators to safely produce a plasma state trajectory leading to and then maintaining the desired experimental conditions. Often new physics understanding leads to new or modified control requirements that use existing actuators in new ways. We describe several important DIII-D PCS design and test tools that support implementation and optimization of algorithms. We describe selected algorithms and the ways they fit within the PCS architecture, which in turn allows great flexibility in designing, constructing, and using the algorithms to reliably produce a desired complex experimental environment. Control algorithms, PCS interfaces, and design and testing tools are described from the perspective of the physics operator (PO), who must operate the PCS to achieve experimental goals and maximize physics productivity of the tokamak. For example, from a POs (and experimental team leader's) standpoint, a PCS algorithm interface that offers maximum actuator, algorithmic, and measurement configuration flexibility is most likely to produce a successful experimental outcome. However, proper constraints that limit flexibility in use of the PCS can also help to maximize effectiveness. For example, device limits and safety must be built into the PCS, sometimes at the algorithm level. We show how the DIII-D PCS toolset enables rapid offline testing of a new or modified algorithm in a simulated tokamak environment. Finally, we illustrate usage of PCS-based checklists and procedures that enhance experimental productivity, and we describe an asynchronous condition detector system within the PCS that enhances device safety and enables complex experiment design.

The KSTAR (Korea Superconducting Tokamak Advanced Researc pursued to develop key technologies for superconducting tokamak operation and to contribute to a few research items for ITER relevant issue...

We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR ∝ [RTAR]2[L||]6/7 and TTAR ∝ [RTAR]−2[L||]−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.

In recent 2 years, various algorithms to control plasma shape, current and density have been implemented or improved for EAST tokamak. These plasma control performances have been verified by either simulated or actual experimental operation, and thus plasma control basis has been established for the long pulse operation and high performance H-mode plasma operation with low hybrid wave (LHW) and ion cyclotron resonance frequency (ICRF) heating. Startup simulation has been done by using TOKSYS code for the plasma breakdown in either 3.1 Wb or 4.5 Wb initial poloidal flux state and the scenarios proved to be robust and used for routine operation. Various shape configurations have been well feedback controlled by using ISOFLUX limited, double-null or single null algorithms based on RTEFIT equilibrium reconstruction. For the long pulse operation, strike point control and magnetics drift compensation have been implemented in the plasma control system (PCS). To improve the operation safety and efficiency, the verification of magnetic diagnostics before plasma breakdown has been demonstrated adequate to prevent a discharge in case of key sensor failure.

In order to improve the plasma control performance and enhance the capability for advanced plasma control, new algorithms such as PEFIT/ISOFLUX plasma shape feedback control, quasi-snowflake plasma shape development and vertical control under new vertical control power supply, have been implemented and experimentally tested and verified in EAST 2014 campaign. P-EFIT is a rewritten version of EFIT aiming at fast real-time equilibrium reconstruction by using GPU for parallelized computation. Successful control using PEFIT/ISOFLUX was established in dedicated experiment. Snowfldivertor plasma shape has the advantage of spreading heat over the divertor target and a quasi-snowflake (QSF) configuration was achieved in discharges with Ip = 0.25 MA and Bt = 1.8T, κ∼1.9, by plasma position feedback control. The shape feedback control to achieve QSF shape has been preliminary implemented by using PEFIT and the initial experimental test has been done. For more robust vertical instability control, the inner coil (IC) and its power supply have been upgraded. A new control algorithm with the combination of Bang-bang and PID controllers has been developed. It is shown that new vertical control power supply together with the new control algorithms results in higher vertical controllability.

The Korea Superconducting Tokamak Advanced Research (KSTAR) device is aimed at advanced tokamak (AT) research. Three years have passed since it achieved its first plasma in 2008. Because it is a superconducting machine and is working toward AT research, it has unique features in terms of the machine engineering and operation. The toroidal field (TF) magnet coils are made of Nb3Sn, which provide high TFs up to 3.5 T, and have been fully tested. The poloidal field (PF) magnet coils, consisting of both Nb3Sn and NbTi, which have a maximum current of 25 kA in their design, were tested up to 15 kA. A thermal hydraulic analysis is being conducted for PF magnet coil operation. All plasma-facing components (PFCs) are equipped with water cooled graphite tiles and have the capability of being baked up to 350°C. A startup scenario, which considered both the effect of the ferromagnetic material in the cable in conduit conductor jacket of the magnet coils as well as a nonferromagnetic up-down asymmetry in the cryostat structure, was developed and demonstrated its effectiveness by the last two year's reliable operations. Passive stabilizers and in-vessel control coils (IVCCs) are key components to realize AT operation in KSTAR. The segmented IVCC coils were connected to form circular coils for internal vertical control in 2010, and diverted plasmas with high elongation (κ~1.8, δ>;0.6) were achieved. A neutral beam injection (NBI) system was developed aiming at 2 MW, 300 s per ion source which meets the long-pulse requirement of KSTAR. An NBI ion source with a power of 1.7 MW at 100 kV has been commissioned for 10 s. Finally, ELMy H-modes were successfully produced with 1.3-MW NBI power at a plasma current of 0.6 MA in the 2010 campaign. The first H-mode discharge (#4200) in KSTAR was achieved one year earlier than officially planned and was done at BT = 2 T with Ip = 0.6 MA in a well-balanced double null configuration after boronization on the PFC. Successful operations in the early days of KSTAR including H-mode experiments revealed the capability of advanced and steady-state operation which is essential for the international thermonuclear experimental reactor (ITER) and future fusion reactors.

We report on recent DIII-D experiments that integrate edge localized mode (ELM) suppression using resonant magnetic perturbations (RMPs) with divertor heat flux reduction under radiating divertor conditions. Our results illustrate the limitations in maintaining ELM suppression at gas puffing levels that were representative of good puff and pump operation in previous experiments without RMP. The electron pressure gradient in the pedestal (∇Pe) increased steadily during gas puffing and ELMs returned once ∇Pe reached values consistent with the peeling–ballooning stability limit, as determined by edge stability analysis. Even with this return of ELMs, a radiating divertor with RMP generated higher levels of total radiated power (∼40%) than comparable standard ELMing discharges without RMP at the same density. Differences in the accumulation of the seed argon in the core plasma between RMP and non-RMP during puff and pump were less than 20%.

The DIII-D digital plasma control system (D3PCS) has been in use since 1993 controlling DIII-D plasmas. Control research and D3PCS development at DIII-D has revealed aspects of advanced tokamak control which can help inform the ITER design. The D3PCS has also been adapted for use at several fusion devices worldwide, which has allowed the DIII-D team to obtain a significant understanding of common requirements for plasma control on multiple fusion devices, along with substantial experience in the alternative computing, data acquisition, and networking technologies presently available. We describe some of what has been learned and highlight the relevance of these lessons for ITER Control, Data Access and Communication (CODAC). Several capabilities of the D3PCS are described in the context of ITER requirements. This description touches on multiple subsystems described in the CODAC conceptual design. We also discuss features of D3PCS architecture that are appropriate for today's experimental devices but may not be appropriate for ITER, which requires a much more comprehensive system for ensuring device safety. Even for these applications, the knowledge gained in implementing methods to aid in ensuring device safety on present devices provides useful guidance for an eventual ITER solution.

Alfven Eigenmodes are destabilized at the DIII-D pedestal during transient beta drops in high poloidal beta discharges with internal transport barriers (ITBs), driven by n=1 external kink modes, leading to energetic particle losses. There are two different scenarios in the thermal beta recovery phase: with bifurcation (two instability branches with different frequencies) or without bifurcation (single instability branch). We use the reduced MHD equations in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles as well as the geodesic acoustic wave dynamics, to study the properties of the instabilities observed in the DIII-D high poloidal beta discharges and identify the conditions to trigger the bifurcation. The simulations suggest that instabilities with lower frequency in the bifurcation case are ballooning modes driven at the plasma pedestal, while the instability branch with higher frequencies are low n (n

Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5× reduction in the peak heat flux for many energy confinement times (2–3 s) without any adverse effects on core plasma performance.

This paper briefly outlines the current status of CODAC system of the EAST tokamak. Integrated central system has been working for synchronizing all the subsystems and responsible for the integral safety protection by interlock. Data acquisition system has been established for the discharge with duration up to 100 seconds. Various data visualization tools such as EASTVIEWER to view the flux surfaces, RTSCOPE to view plasma boundary and CCD image in real-time, WEBSCOPE to view the diagnostic data via a web browser and EASTVOD to view and search plasma discharge image at different time, haven developed for EAST operation. Plasma control system is inherited from DIII-D, and has been matured to a stage to feedback control the EAST plasma shape, density and current in real time.

Recent commissioning of two major fully superconducting (SC)-shaped tokamaks, Experimental Advanced Superconducting Tokamak (EAST) and Korean Superconducting Tokamak Advanced Research (KSTAR), represents a significant advance in magnetic fusion research. The key to commissioning success in these complex and unique tokamaks was as follows: 1) use of a robust, flexible plasma control system (PCS) based on the validated DIII-D design; 2) use of the TokSys design and modeling environment, which is tightly coupled with the DIII-D PCS architecture for first-plasma scenario development and plasma diagnosis; and 3) collaborations with experienced internationally recognized teams of tokamak operations and control experts. We provide an overview of the generic modeling environment and plasma control tools developed and validated within the DIII-D experimental program and applied through an international collaborative program to successfully address the unique constraints associated with the startup of these next-generation tokamaks. The unique characteristics of each tokamak and the machine constraints that must be included in device modeling and simulation, such as SC coil current slew rate limits and the presence of nonlinear magnetic materials, are discussed, along with commissioning and initial operational results. Lessons learned from the startup experience in these devices are summarized, with special emphasis on ramifications for International Thermonuclear Experimental Reactor (ITER).

The DIII-D device began operation in 1986, and a fully digital plasma control system (PCS [J.L. Luxon, Nucl. Fusion 42, 614 (2002).]) was implemented in 1993. Over time, the success of the PCS to exploit the inherent versatility of the DIII-D device led to a philosophy of using the PCS to control all available plasma system actuators. This made the PCS a very powerful physics tool that is at the core of Physics Operations at DIII-D. The complexity of the DIII-D device and all the systems the PCS must control makes proper setup of the PCS for new experiments a daunting task. A cadre of physicists specially trained in PCS operation forms the Physics Operations staff at DIII-D. They are the interface between experimental plans and successful execution and, as such, are a critical component of each experiment. Physics Operations is also a set of tools and procedures. We will briefly examine some of those tools, such as the TokSys control design and modeling environment, and the ‘smart’ PCS setup checklist, that greatly reduces human error in reconfiguring the PCS for a new experiment. We will examine the procedures that allow efficient use of those tools and some of the human factors that can affect productivity.