Hyatt A.,General Atomics |
Humphreys D.A.,General Atomics |
Welander A.,General Atomics |
Eidietis N.,General Atomics |
And 9 more authors.
IEEE Transactions on Plasma Science | Year: 2014
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. © 1973-2012 IEEE.
The team is led by Dr. Xianzu Gong of ASIPP and Dr. Andrea Garofalo of General Atomics (GA) in San Diego. Using both China's EAST facility and the DIII-D National Fusion Facility, operated by GA for the U.S. Department of Energy, the team has investigated the "high-bootstrap current" scenario, which enhances self-generated ("bootstrap") electrical current to find an optimal tokamak configuration for fusion energy production. Magnetic fusion energy research uses magnetic fields to confine plasma (ionized gas) heated to temperatures hotter than the Sun's core. This enables the ions to fuse and release excess energy that can be turned into electricity, harnessing the Sun's power on Earth. The most developed configuration is the tokamak, and the team's work helps prepare for the 500-megawatt ITER fusion research facility that is currently being built in France by a consortium of 35 nations, including China and the U.S. This joint U.S.-China experiment directly demonstrates the stabilizing effect of reducing the plasma-wall distance in tokamaks with high plasma pressure and large bootstrap current fraction, according to Dr. Gong, who said, "I think, in simple terms, these experiments may provide better physics and operation foundation for ITER plasmas." The focus was on resolving the "kink mode" instability, a wobbling effect that reduces performance, by moving the plasma closer to the vessel's wall, Dr. Garofalo explained . Operating closer to the wall suppresses the kink mode and enables higher pressure inside the tokamak, the toroidal or doughnut-shaped steel-lined fusion device. This gives rise to "pressure-driven" plasma flows that maintain the confinement quality even with lower external injection of velocity. "This is unlike any other regime," said Dr. Garofalo. "It's very risky to move the plasma that close to the wall. The chief operator said 'You can't do that anymore, you're going to damage the machine,' so it was a struggle to prove our theory was correct." The gambit paid off. Moving the plasma closer to the wall removed the kink mode and enabled higher plasma pressure, which, in turn, makes the plasma less dependent on externally injected flow. This is important because in a tokamak reactor, such as ITER, it is very difficult and expensive to drive a rapid plasma flow with external means. The team performed the most recent bootstrap exploration in DIII-D, following-up work on the record-setting milestone achieved at China's EAST tokamak, where GA scientists have also been collaborating. An ASIPP scientist Dr. Qilong Ren will deliver the invited talk on the topic of Magnetic Confinement-Experiments. While fusion has been in the public domain since the 1950s and its advances have been achieved by teams around the world, this U.S.-China team is setting new milestones in global cooperation. For realization of magnetic fusion energy, global cooperation is needed, said Dr. Gong of ASIPP, who cited the EAST/DIII-D partnership as "an efficient and effective new model" for international science collaborations that benefits both partners and the field of study. "We have made a very good start of international collaboration in fusion research between China and the U.S., and we are very proud to be a pioneer in this field," said Dr. Gong. More information: Abstract: KI2.00004 Progress Toward Steady State Tokamak Operation: Exploiting the high bootstrap current fraction regime Session Session KI2: MFE Regime Optimization
Pautasso G.,Max Planck Institute for Plasma Physics (Garching) |
Zhang Y.,ASIPP |
Reiter B.,Max Planck Institute for Plasma Physics (Garching) |
Giannone L.,Max Planck Institute for Plasma Physics (Garching) |
And 11 more authors.
Nuclear Fusion | Year: 2011
This paper describes the most recent contributions of ASDEX Upgrade to ITER in the field of disruption studies. (1) The ITER specifications for the halo current magnitude are based on data collected from several tokamaks and summarized in the plot of the toroidal peaking factor versus the maximum halo current fraction. Even if the maximum halo current in ASDEX Upgrade reaches 50% of the plasma current, the duration of this maximum lasts a fraction of a ms. (2) Long-lasting asymmetries of the halo current are rare and do not give rise to a large asymmetric component of the mechanical forces on the machine. Differently from JET, these asymmetries are neither locked nor exhibit a stationary harmonic structure. (3) Recent work on disruption prediction has concentrated on the search for a simple function of the most relevant plasma parameters, which is able to discriminate between the safe and pre-disruption phases of a discharge. For this purpose, the disruptions of the last four years have been classified into groups and then discriminant analysis is used to select the most significant variables and to derive the discriminant function. (4) The attainment of the critical density for the collisional suppression of the runaway electrons seems to be technically and physically possible on our medium size tokamak. The CO2 interferometer and the AXUV diagnostic provide information on the highly 3D impurity transport process during the whole plasma quench. © 2011 IAEA, Vienna.
Schaffer M.J.,General Atomics |
Snipes J.A.,ITER Organization |
Gohil P.,General Atomics |
De Vries P.,EURATOM |
And 37 more authors.
Nuclear Fusion | Year: 2011
Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L-H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δv/v ∼ 60% via non-resonant braking. Changes to global Δn/n, Δβ/β and ΔH 98/H98 were ∼3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small. © 2011 IAEA, Vienna.
Ballarino A.,CERN |
Bauer P.,ITER Organization |
Bi Y.,ASIPP |
Devred A.,ITER Organization |
And 9 more authors.
IEEE Transactions on Applied Superconductivity | Year: 2012
Following the design, fabrication and test of a series of trial leads, designs of the three types of current leads required for ITER have been developed, and targeted trials of specific features are in progress on the way to fabrication and testing of prototype units. These leads are of the hybrid type with a cold section based on the use of high temperature superconductor (HTS) and a resistive section cooled by forced flow of helium gas, optimized for operation at 68 kA, 55 kA and 10 kA. The leads incorporate relevant features of the large series of current leads developed and constructed for the CERN-LHC, relevant features of the trial leads built for ITER, and additional features required to fully satisfy the exigent constraints of ITER with regard to cooling, insulation, and interfaces to feeder and powering systems. In this report a description of the design of the leads is presented, together with plans for the preparation of prototype manufacture and testing at ASIPP. © 2011 IEEE.