J.T. Scoville1, B.J. Crowley1, B.A. Grierson2, M. Hansink1, A. Nagy2, J.M. Rauch1, 1General Atomics, PO Box 85608, San Diego, California 92186, USA
2Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451, USA
The Neutral Beam Injection System (NBI) on the DIII-D tokamak includes four beamlines, each equipped with two ion sources. The eight ion sources operate nominally at 75-80 kV and are capable of injecting up to ~20 MW for plasma heating and current drive. One of the beamlines was modified several years ago to allow for off-axis injection. Recently, another beamline was modified to provide additional off-axis beams that can be aimed toroidally in either the co-injection or counter-injection direction. The high-precision movement system that was installed in this new beamline configuration provides off-axis injection at an angle of 18.4° above horizontal and slewing 40° toroidally between the co- and counter-injection positions. This upgrade provides unprecedented flexibility in the DIII-D neutral beam system, allowing up to ~10 MW of off-axis beams with balanced injection (zero torque) or ~20 MW of beams with co-injection (including ~10 MW off-axis). The result is a wider parameter space for DIII-D, aiding in the study of high beta, low rotation, steady-state advanced tokamak scenarios that are relevant to the ITER device.
A description of the new co/counter off-axis neutral beam (CCOANB) system will be presented, along with a discussion of some of the installation and commissioning issues for the beamline. Results will be reported from the first physics experiments that utilized the flexible new CCOANB system.
*This work supported in part by the US DOE under DE-FC02-04ER54698.
Contact: Tim Scoville, firstname.lastname@example.org
Avoidance of vertical displacement events in DIII-D using a neural network growth rate estimator B.S. Sammuli, J.L. Barr, D.A. Humphreys
DIII-D National Fusion Facility, General Atomics, PO Box 85608, San Diego, CA, USA
Robust disruption avoidance techniques are critical for the development of reliable fusion reactor devices. A viable reactor will require non-disruptive, long pulse operation where simply shutting down a discharge is undesirable. To achieve such performance, the plasma must be controlled to continuously avoid hazardous regimes instead of asynchronously aborting. A recent experiment on DIII-D demonstrated for the first time real-time control of proximity to a disruptive instability boundary. In particular, the vertical growth rate was regulated so as not to exceed DIII-D’s vertical controllability limit. The open-loop growth rate was estimated in real time on the DIII-D plasma control system using a neural network model trained with tens of thousands of DIII-D shots. The model was trained to replicate the results of RZRIG, a rigid displacement code for calculating the growth rate. Once trained, producing an estimate using the neural network is multiple orders of magnitude faster than RZRIG, thereby making the calculation suitable for real-time execution. The control system regulated the estimated growth rate by adjusting plasma elongation and inner-gap distance, and this regulation was shown to reliably avoid vertical displacement event disruptions of the plasma. This work presents these experimental results, including the dynamic performance and the effectiveness of the control technique. Details are presented on the training of the neural network model, including concerns such as hyperparameter tuning and uncertainty quantification. Additionally, the methodology for embedding the neural network into the control system is discussed.
Work supported by General Atomics’ Internal Research and Development Funding and in part by the U.S. 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 Award No. DE- FC02-04ER54698.
Contact: Brian Sammuli, email@example.com
Real-time estimation and feedback control of surface heat flux on the DIII-D tokamak
H. Anand1, D. Eldon1, D. Humphreys1, J. Boedo2, Al. Hyatt1, B. Sammuli1, F. Scotti1, A. Welander1, and J. Barr1 1DIII-D National Fusion Facility, General Atomics, PO Box 85608, San Diego, CA 92186, USA 2University of California, San Diego Center for Energy Research, MS 0417 La Jolla, CA 92093, USA
Future tokamaks will require robust technologies for the mitigation of heat exhaust onto the plasma-facing components (PFCs). A feedback system has been developed at DIII- D that estimates and controls in real time (RT) the heat flux to the PFCs. The system utilizes a control-oriented model, based on real-time equilibrium reconstruction (RTEFIT), which is then used to describe the deposited heat flux as a poloidal flux function with user- specified parameters for the power exhausted into the scrape-off layer (SOL) and the SOL heat flux width. The peak heat flux density on the PFCs is then regulated with the poloidal field coils by changing the poloidal flux expansion and thereby the plasma-wetted area at the divertor target. Real-time estimation of the peak power flux from the model- based approach is compared with off-line infra-red measurements for various DIII-D plasma discharges. A nonlinear free-boundary simulation code (GSevolve) that evolves the Grad-Shafranov equilibrium including current and pressure profiles is used for simulating the closed loop response and for determining the control parameters. The implementation and first experimental results of application during a DIII-D plasma campaign are reported.
Acknowledgment: This work was supported in part by the US Department of Energy under grant DE-FC02-04ER54698
Contact: Himank Anand, firstname.lastname@example.org
The Impact of Beam Optics and Gas Dynamics on the Performance Limitations of the DIII-D Off-Axis Neutral Beams.*
B.J. Crowley, D.Glibert, J.M. Rauch and J.T. Scoville General Atomics, PO Box 85608, San Diego, California 92186, USA
The Neutral Beam Injection System (NBI) on the DIII-D tokamak includes four beamlines, each equipped with two ion sources. The eight ion sources operate nominally at 75-80 kV and are capable of injecting up to ~20 MW for plasma heating and current drive. Two of the beamlines have been modified in the past to introduce the capability of off-axis neutral beam current drive. A necessary consequence of the modification is the effective constriction of the duct that connects the beamline to the tokamak vacuum vessel. The constricted duct is now the limiting component of the beamline in terms of its ability to handle beam power. The duct is subject to two types of power interception, namely direct interception from particles at the edges of the beam and interception from re-ionized particles whose trajectories are bent and focused by the high magnetic fields in the duct region. Options for mitigating the limitations include improving the focusing and collimation of the beam. In addition, a better understanding of gas dynamics may allow optimizing the gas flow in order to minimize or eliminate the reionized beam power. This paper reports on a modeling and experimental study that includes finite element electrostatic modeling of the ion optics, Monte Carlo simulations of gas dynamics and Runge-Kutta based particle tracing in electric and magnetic fields. The model results are used to determine the optimal settings of high voltage power supplies and gas flows that minimize duct interaction and maximize power and energy input to the DIII-D tokamak.
*This work supported in part by the US DOE under DE-FC02-04ER54698.
Contact: Brendan J. Crowley, email@example.com