A Handheld, Free Roaming, Data Display for DIII-D Diagnostic Data

Standard handheld test instruments such as voltmeters and portable oscilloscopes are useful for making basic measurements necessary for the operation and maintenance of large experiments such as the DIII-D magnetic fusion research facility. Some critical diagnostic information, however, is available only on system computers. Often this diagnostic information is located in computer databases and requires synthesis via computational algorithms to be of practical use to the technician. Unfortunately, this means the data is typically only available via computer screens located at fixed locations. One common way to provide mobile information is to have one operator sit at a console and read the data to the mobile technician via radio. This is inefficient in as much as it requires two people. Even more importantly the operator-to-technician voice link introduces significant delays and errors that may hinder response times. To address these concerns of personnel utilization and efficiency, we have developed a remote display based on an rf-data link that can be carried with a technician as he moves about the facility. This display can provide the technician with any information needed from the stationary database. This paper will discuss the overall architecture as well as the individual modules for the mobile data display. Lessons learned, as well as techniques for improving the usefulness of such systems, will be presented

Overview of DIII-D off-axis neutral beam project

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. © 2011 IEEE

DIII-D research towards establishing the scientific basis for future fusion reactors

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (similar to 30 kPa) and stored energy (3.2 MJ) with H-98y2 approximate to 1.6-2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric beta(TE) independent of current between q(95) = 2.8-3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses approximate to 0 injected torque and the operating space is more ITER-relevant. Finally, the high-beta(N) (<= 3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving similar to 40% divertor heat flux reduction using either argon or neon with P-tot up to 15 MW

Overview of recent experimental results from the DIII-D advanced tokamak programme

The goals of DIII-D advanced tokamak (AT) experiments are investigation and optimization of the upper limits of energy confinement and MHD stability in a tokamak plasma, and simultaneous maximization of the fraction of non-inductive current drive. Significant overall progress has been made in the past two years, as the performance figure of merit βNH89P of 9 has been achieved in ELMing H mode for over 16τE without sawteeth. The tokamak was also operated at βNH ≈ 7 for over 35τE or 3τR, with the duration limited by the hardware. Real time feedback control of β (at 95% of the stability boundary), optimizing the plasma shape (e.g., δ, divertor strike and X points, double/single null balance) and particle control (ne/nGW ≈ 0.3, Zeff &lt; 2.0) were necessary for the long pulse results. A new quiescent double barrier (QDB) regime with simultaneous inner and edge transport barriers and no ELMs has been discovered with a βNH89P of 7. The QDB regime has been obtained to date only with counter NBI. Further modification and control of internal transport barriers (ITBs) has also been demonstrated with impurity injection (broader barrier), pellets and ECH (strong electron barrier). The new Divertor-2000, a key ingredient in all these discharges, provides effective density, impurity and heat flux control in the high triangularity plasma shapes. Discharges at ne/nGW ≈ 1.4 have been obtained with gas puffing by maintaining the edge pedestal pressure; this operation is easier with Divertor-2000. We are developing several other tools required for AT operation, including real time feedback control of resistive wall modes with external coils and control of neoclassical tearing modes with ECCD

DIII-D research advancing the physics basis for optimizing the tokamak approach to fusion energy

Funding Information: This material is based upon work supported by the US Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-FC02-04ER54698 and DE-AC52-07NA27344. Publisher Copyright: © 2022 IAEA, Vienna.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.Peer reviewe