The General Atomics Tokamak Model (GATM) is a recently developed finite element (FE) environment for assessing the performance of tokamak magnet-plasma systems. Its generic environment is designed to assess magnet performance parameters for existing tokamaks, such as Doublet-III-D (DIII-D) and next-generation tokamak. GATM’s environment consists of a number of models, including a Comsol 3-D FE model and routines connecting this model to other GA tools. GATM consists of a 2-D axisymmetric module characterizing Poloidal Fields (PFs) system magnetic properties, including an equilibrium fitting code (EFIT) plasma representation. The 2-D module contains a central solenoid (CS) structural model for stress and deflection assessment. Toroidal field (TF) coils are characterized by a 3-D magnetics module for TF calculations and simulation of TF nonaxisymmetric ripple. PF and TF magnetics models provide input to a 3-D TF structural module, which simulates both in-plane and out-of-plane stress. A 3-D TF center post module uses input from the other three modules to evaluate the performance of a superconducting coil including details of the case, winding-pack, and superconducting cable. Model geometry, (Plasma, TF, CS, PF, and structures) is fully parametrized and allows for rapid assessment of arbitrary tokamak configurations. The GATM model is under development and is intended to interface with other general atomics models including: GA System Code (GASC), dynamic tokamak system model (TokSys), recently developed FUsion Synthesis Engine (FUSE), and DIII-D database, including EFIT. Models and current states generated in the GATM environment are output to engineering relevant codes ANSYS and SolidWorks. The article provides an overview of the present GATM environment and its application to DIII-D’s TF coil and an advanced tokamak (AT) Fusion Pilot Plant (FPP).

DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter- I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high- Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L–H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼ 8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co- I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.

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 x del 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 less than or similar to 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 x del 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.

In the DIII-D tokamak, one of the most powerful techniques to control the density, temperature and plasma rotation is by eight independently modulated neutral beam sources with a total power of 20 MW. The rapid modulation requires a high degree of reproducibility and precise control of the ion source plasma and beam acceleration voltage. Recent changes have been made to the controls to provide a new capability to smoothly vary the beam current and beam voltage during a discharge, while maintaining the modulation capability. The ion source plasma inside the arc chamber is controlled through feedback from the Langmuir probes measuring plasma density near the extraction end. To provide the new capability, the plasma control system (PCS) has been enabled to change the Langmuir probe set point and the beam voltage set point in real time. When the PCS varies the Langmuir set point, the plasma density is directly controlled in the arc chamber, thus changing the beam current (perveance) and power going into the tokamak. Alternately, the PCS can sweep the beam voltage set point by 20 kV or more and adjust the Langmuir probe setting to match, keeping the perveance constant and beam divergence at a minimum. This changes the beam power and average neutral particle energy, which changes deposition in the tokamak plasma. The ion separating magnetic field must accurately match the beam voltage to protect the beam line. To do this, the magnet current control accurately tracks the beam voltage set point. These new capabilities allow continuous in-shot variation of neutral beam ion energy to complement the discontinuous "on or off" modulation method presently used to control average beam power and torque input. (C) 2017 Elsevier B.V. All rights reserved.

Recently, an Edge Localized Mode (ELM) control coil was developed for use on the DIII-D tokamak. The coil design represented a significant challenge due primarily to the requirement for the coil insulation to withstand bakeout temperatures of 400°C for extended periods. This requirement ruled out most common organic insulating systems and necessitated a significant prototyping and development effort, leading to the selection of an advanced high temperature glass/polyimide resin system. The development included developing a heating mechanism that provided the discrete temperature ramp cycles and cure cycles required by this exotic resin. To complicate matters, the resin had a limited shelf life. Additionally the coil was small and rectangular in shape with rather small corner radii. This created a corner buildup that was not previously encountered and made dimensional control difficult. Another unique design requirement was the need to apply a sufficient internal pre-load to the wound and cured coil to insure there will be no relative motion between the coil and the Inconel case due to Lorentz forces from the 4 Tesla toroidal field on the vessel center post. This led to development of very unique leaf springs and a significant research and development effort coupled with an equally arduous finite element analysis effort. A satisfactory prototype was produced. This paper will focus primarily on the manufacturing challenges and discuss the prototyping effort.

DIII-D has four neutral beamlines (NB). Each of these beamlines has two ion sources, each of which injects up to 2.5 MW for 3 s. These beamlines intersect the vacuum vessel at an angle of 19.5 deg off from radial, enabling current drive in the same direction as the plasma current (co-injection). In 2004, one of these beamlines (210 deg) was rotated to provide counter-injection (opposite of plasma current). A different beamline (150 deg) has been modified to have the capability to provide off-axis neutral beam current drive. The goal of the off-axis injection is to have the center of the ion sources aimed at a position 40 cm below the geometric center of the plasma. To achieve this off-axis injection, the beamline requires a mechanical lifting system that can elevate the beamline up to 16.5 deg from horizontal. The beamline also requires more strongly vertically focused ion sources (in order to pass the beam through a reduced effective aperture) as well as modified internal components. Additionally, the design of the new internal components incorporated modifications to allow for the doubling of ion source pulse lengths without the need for active cooling. This paper discusses the various beamline system design requirements for off-axis injection, as well as the results from the actual commissioning of the beamline. Overviews of the design and performance of mechanical lifting system (hydraulics and controls), focused ion sources, flexible beamline support systems (vacuum, cryogenic, power and water cooling), and internal beamline collimators are included. Additionally, the in-vessel monitoring and shine-through protection requirements are discussed. The actual data obtained during beamline commissioning and during normal physics operations is also presented.

Prompt runaway electron bursts, generated by rapidly cooling DIII-D plasmas with argon “killer” pellets, are used to test a recent knock-on avalanche theory describing the growth of multiMeV runaway electron currents during disruptions in tokamaks. Runaway current amplitudes, observed during some but not all DIII-D current quenches, are consistent with growth rates predicted by the theory assuming a pre-current quench runaway electron density of approximately 1015 m-3. Argon “killer” pellet modeling yields runaway densities of between 1015-1016 m-3 in these discharges. Although knock-on avalanching appears to agree rather well with the measurements, relatively small avalanche amplification factors combined with uncertainties in the spatial distribution of pellet mass and cooling rates make it diEcult to unambiguously confirm the proposed theory with existing data. Additional measurements are proposed which should enable us to definitively test the theory.

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.

Major upgrades to the DIII–D facility have been performed that significantly enhance the capability of both the DIII–D device and the entire facility. The most significant of these include the rotation of a neutral beam line, installation of a new lower divertor, and a significant set of new and enhanced diagnostics. The upgrades and initial results are presented in this paper.

Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. In 2000, these coils also demonstrated benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modeling has shown that substantial performance improvements could be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and were power tested successfully after several bakes to 350/spl deg/C. A full set of twelve internal coils and related magnetic sensors are now operational in the DIII-D tokamak. The design requirements for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator. Elastic-plastic analysis was used to demonstrate acceptable thermal stresses during baking conditions. Analysis determined the optimum water cooling channel diameter. The coils were tested in high toroidal field to the limit of the power supply of 4.5 kA DC with inductance-limited current for frequencies between 300 Hz and 1000 Hz. Recent results are presented.

The installation of new internal magnetic coils (I-Coils) in the DIII-D tokamak at General Atomics required extensive additions to the experiment data acquisition and protection capabilities. This set of 12 coils (up to 7 kA each) is designed to allow improved feedback stabilization of resistive wall modes which limit the plasma performance. The acquisition and signal conditioning needs of the I-Coil power system presented an opportunity to try a new data acquisition approach which increased both the sampling rate and sample size per channel compared to the standard DIII-D CAMAC acquisition equipment. A 96 channel compact-PCI (cPCI) digitizer system was purchased for the I-Coil project to acquire up to approximately 380 MB of power supply and coil current data per plasma discharge. Additional instrumentation and control was provided to protect personnel, the new coils, the tokamak, the facility and improve machine availability. This paper will present discussions of technical and programmatic requirements, bases for requirements, the design selection outcome, installation experience, integration issues, commissioning experience, and lessons learned. The data acquisition system is described in detail including a conservative signal isolation scheme, signal grounding standards, anti-aliasing filters, and synchronization of acquisition. Protection interlocks are described, including high voltage isolation, water flow measurement, and the coil grounding-shorting switches.

Halo currents, which flow in plasma residing on open field lines during tokamak disruptions, can apply large, highly localized forces to plasma facing components. Successful models of halo and core plasma current evolution developed in the last decade account for many of the processes, and allow prediction of halo currents provided the post-thermal quench (TQ) core and halo resistivities and halo width evolution are given.1 Use of the KPRAD impurity radiation code with these models allows reasonably accurate prediction of post-TQ average plasma resistivity when the ohmic input power is balanced by radiation from a known quantity of identified impurity species.2 Key elements lacking predictability remain the halo width evolution, effectiveness of impurity penetration, resulting impurity species concentrations in the post-TQ plasma, and separate evolution of the resistivities of the core and halo. We describe a predictive model for evolution of the key relevant characteristics of a plasma halo in DIII-D disruptions, including halo width and resistivities. The effective halo width is set by plasma motion and current diffusion across open field lines in the post-TQ plasma, typically filling the entire machine aperture with current by the end of the current quench. Integration of KPRAD with dynamic plasma and field line evolution models allows core and halo resistivities to be determined separately. Although the amounts of impurity sputtering, ablation, and penetration into mitigated or naturally disrupting plasmas are not yet predictable, sufficient simulation detail is now available to allow realistic scoping studies based on varying the assumed quantity of first wall material or actively injected impurity in the core plasma. The model is applied to DIII-D experiments and ITER disruption scenarios.

The contributions of the DIII-D tokamak toward the understanding and control of disruptions are reviewed. Disruptions are found to be deterministic, and the underlying causes of disruption can therefore be predicted and avoided. With sufficiently rapid detection, possible damage from disruptions can be mitigated using an understanding of disruption phenomenology and plasma physics. Regimes of high beta are readily available in DIII-D and provide access to relatively high energy density disruptions, despite DIII-D's moderate magnetic field and size. DIII-D, with all-graphite wall armor and wall conditioning between discharges, has proven highly resilient to the deleterious effects that disruptions can have on plasma operations. Simultaneously, exploitation and adaptation of DIII-D's extensive core and edge plasma diagnostic set have allowed for unique plasma measurements during disruptions. These measurements have tied into the development of several physical models used to understand aspects of disruptions, such as magnetohydrodynamic growth at the disruption onset, radiation energy balance through the thermal quench, and halo currents during the current quench. Based on this fundamental understanding, DIII-D has developed techniques to mitigate the harmful effects of disruptions by radiative dissipation of the plasma energy and extrapolated these techniques for possible use on larger devices like ITER.

The advancement of plasma control techniques has enabled significant progress to be made toward the scientific understanding and realization of Advanced Tokamak operation on DIII-D. The Advanced Tokamak features fully non-inductive current drive, operation at high plasma pressure and high energy confinement time. These features require efficient current drive systems, simultaneous control of plasma current and pressure profiles, and active feedback control of plasma instabilities. A number of key systems on DIII-D have been developed to provide this control capability. A versatile electron cyclotron heating and current drive system is routinely providing in excess of 2 MW of power for pulse lengths from 2 to 5 s. This system has been used to provide off-axis current drive, direct electron heating and pressure profile modification, and stabilization of the Neoclassical Tearing Mode instability. A combination of control of magnetic error fields, neutral beam induced plasma rotation, and active feedback stabilization using both external and internal non-axisymmetric coil systems has been used to stabilize the Resistive Wall Mode at high values of plasma pressure. Control of the ELM instability has recently been demonstrated using the newly installed internal coil system. The higher speed and expanded real-time diagnostic capability of our recently upgraded plasma control system permits these various control techniques to be simultaneously integrated to achieve our high performance discharges. This has resulted in fully non-inductively driven plasmas with beta(N) = 3.5 and beta(T) = 3.6% sustained for up to 1 s. Upgrades and facility modifications to further enhance our control and scientific capabilities including rotation of a neutral beamline, expanded EC system power, and installation of a new lower divertor are discussed.

For most of 2003, DIII-D operated 12 new in-vessel outer wall mounted control coils. The single turn, rectangular coils are mounted in 2 levels of 6 coils each. The coils were used for many experiments such as suppression of the resistive wall mode, for correction of magnetic field imperfections and for creation of an ergodic edge magnetic field for the suppression of edge localized modes. During operation with a maximum current of 4.5 kA at 100 Hz, one coil developed a leak through the stainless barrier that separates the nitrogen blanketed insulated conductor from vessel vacuum. A pair of coils was taken out of operation for the last month of the year. This paper describes the failure investigation, design, analysis, component testing, repair, system testing and new interlocks for the system that will see significant use in 2004. The crack in the stainless barrier was attributed to low cyclic fatigue related to operation of the coil at 100 Hz, a frequency near the vertical natural frequency of the coaxial lead. Finite element analysis (FEA) after the failure showed that electromagnetic forces on the single conductor section were sufficient to excite the coaxial lead in a vertical mode. The crack likely developed in less than a second. Repair options were limited. The coils are mounted to the walls and removal of PF tiles was discouraged in order to minimize the repair time. Welding on the coil was limited in order to protect the internal polyamide insulator from overheating. Repairs included: 1) seal the leak in the faulted coil, 2) increase the stiffness of the single conductors near the coaxial transition and 3) significantly increase the first natural frequency of the coaxial leads to allow operation to 1000 Hz. In-vessel vibration testing was done at each stage of repair to compare the natural frequency of the three types of leads with that determined by FEA models. Verification testing was done prior to vessel closure. The test included temporary installation of field tolerant strain gages to monitor strain in the stainless for comparison with model results. Permanent vacuum vessel port deflection monitors were added with the hope that excessive lead vibrations could be detected by port deflections for interlock purposes. All 12 coils were successfully repaired, upgraded and test results are encouraging.

Silicon absolute extreme ultraviolet (AXUV) photodiodes have been employed in a disruption radiometer diagnostic for measurement of radiant power in the DIII-D tokamak with a 170 kHz bandwidth. This is motivated by a need to improve the understanding of radiative processes in tokamak disruptions. The diagnostic described in this article has a single line of sight though the central plasma. Accounting for the photon energy dependence of the AXUV photodiode responsivity is made possible by optical filtering, with the aid of spectra from an extreme ultraviolet survey spectrometer. The appropriate effective responsivity for interpretation of the data is lower than the nominal value typically used for the detector. In the current quench phase of disruptions, it is less than half the nominal value. Comparisons with results from a foil bolometer find good agreement.