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Humphreys, D. A.

Publications and source records attributed to Humphreys, D. A..

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Regulation of the central safety factor and normalized beta under low NBI torque in DIII-D

An algorithm has been designed to simultaneously control the central safety factor (q 0 ) and normalized beta (β N ) while ensuring near-zero torque from the neutral beam injection in DIII-D. Feedback control of q 0 and β N in tokamaks can be beneficial due to the close relationship that these variables have with plasma performance and magneto-hydrodynamic stability. In addition, low neutral-beam-torque conditions are of special interest in present devices because future burning-plasma tokamaks such as ITER will most likely operate at very low plasma rotation. The control synthesis of the algorithm presented in this work is based on a linearized, one-dimensional (1D) model of the current-profile dynamics coupled with a zero-dimensional (0D) plasma-energy balance. The actuators considered are neutral beam injection and electron-cyclotron heating and current drive, and discrete logic determines the neutral-beam injection powers that deliver near-zero torque. Here, the algorithm has been tested in nonlinear, 1D simulations using COTSIM (Control-Oriented Transport SIMulator) and in DIII-D experiments, demonstrating satisfactory performance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Exploration of RMP ELM control on ITER-similar shape (ISS) in KSTAR

Recent control advances at KSTAR enabled us to not only establish but also to sustain the ITER-similar shape (ISS) in a stationary manner, matching the ITER-like dimensionless shape parameters in lower single null configuration with vastly contrasting upper/lower triangularities. Here, the kinetic parameters of the ISS are different from the typical KSTAR configurations seen in ELM-RMP suppression experiments. We experimentally demonstrate the ISS-compatible resonant magnetic perturbation (RMP) control in KSTAR for the first time, using the n = 2, +90 degrees phasing RMP.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Observation of fully detached divertor integrated with improved core confinement for tokamak fusion plasmas

Integration of divertor detachment with a high-performance (β N ~ 3, β p > 2, H 98 ~ 1.5) core plasma has been demonstrated in DIII-D high-β p (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-β p plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te < 5 eV), low particle flux, and low heat flux across the entire divertor target plate were obtained by using nitrogen seeding. This detached high-β p plasma is compatible with a newly developed detachment control system which can help optimize the nitrogen gas flow rate. Several features, i.e., the high edge safety factor in the high-β p scenario, impurity injection, closed divertor and reduced heating power requirement due to the high confinement, facilitate the achievement of full divertor detachment at lower density. Instead of degrading global performance, the divertor detachment facilitates the access to an even stronger ITB at large radius with a relatively weak ETB through self-organized synergy between ITB and ETB, leading to sustained high confinement. The strengthening of the large-radius ITB compensates for the ETB degradation associated with divertor detachment. In addition, a weak ETB naturally has smaller edge localized modes (ELMs). In particular, with neon injection, a long-period no-ELM H-mode phase has been achieved simultaneously with high-performance core and partially detached divertor plasmas. Furthermore, these results demonstrate the possibility of integrating excellent core plasma performance with an effective divertor solution, an essential step toward steady-state operation of reactor-grade plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Integration of full divertor detachment with improved core confinement for tokamak fusion plasmas

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 (β N ~ 3, H 98 ~ 1.5) core plasma, using high-β p (poloidal beta, β p > 2) scenario characterized by a sustained core internal transport barrier (ITB) and a modest edge transport barrier (ETB) in DIII-D tokamak. The high-β p 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

High precision strike point control to support experiments in the DIII-D small angle slot divertor

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. Here, 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

First observation of plasma healing via helical equilibrium in tokamak disruptions

A prompt transition of equilibrium states in a disruptive plasma is observed in the presence of three-dimensional (3D) resonant magnetic fields. A dominantly stochastic field line state heals into a 3D equilibrium with intact flux surfaces, resembling a single helicity magnetic island in the plasma core. As a consequence, the current quench process is bifurcated, i.e., from a completed quench accompanied by a vertically unstable disruption to a rapid termination of current decay followed by a full recovery to its predisruption value. Finally, this observation implies prompt stochastic fields healing through the rapid non-linear growth of magnetic island in plasma core may provide a plausible path for a soft-landing scenario of tokamak disruption in high plasma current regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Feedback control of stored energy and rotation with variable beam energy and perveance on DIII-D

Design and testing of a new method for real-time control on DIII-D using the new voltage and perveance variation capability of DIII-D's neutral beam injection system is presented. The approach enables control of power, torque, stored energy, and rotation without requiring pulse-width modulation of the beams. Such modulations are perturbative to experiments, can cause beam reliability problems, and could potentially damage or fatigue accelerator grids, especially looking forward to active control of long-pulse devices and reactors. Here, the control algorithm accounts for beam line loss mechanisms and includes approaches that address the complexity of having three manipulated variables for each of the eight beams and compensate for the rate and magnitude limits affecting the voltage and perveance actuators. Results are presented that demonstrate the performance of the approach in the first feedback control experiments to use the expanded controllability of the DIII-D beams.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Improved fast vertical control in KSTAR

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. Lastly the noticeable improvement in plasma control for plasmas near the vertical stability limit will be discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Disruption Physics and Mitigation on DIII-D

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 β 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. Finally, 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Nonrigid, Linear Plasma Response Model Based on Perturbed Equilibria for Axisymmetric Tokamak Control Design

Tokamak control design relies on an accurate linear model of the plasma response, which can often dominate the local field variations in regions under active feedback control. For example, when fluxes at selected points on the plasma boundary are regulated in DIII–D [1], the plasma response to a change in a coil current gives rise to a flux change which can be larger than and opposite to the flux change caused by the coil alone. In the past, rigid plasma models have been used for linear stability and shape control design. In a rigid model, the plasma current profile is considered fixed and moves rigidly in response to control coils to maintain radial and vertical force balance. In a nonrigid model, however, changes in the plasma shape and current profile are taken into account. Such models are expected to be important for future advanced tokamak control design. The present work describes development of a nonrigid plasma response model for high-accuracy multivariable control design and provides comparisons of model predictions against DIII-D experimental data. Finally, the linear perturbed plasma response model is calculated rapidly from an existing equilibrium solution.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Nonlinear Methods for Current Limit Constraint Satisfaction in Tokamak Plasma Shape Control

The DIII-D tokamak is capable of supporting a wide variety of plasma equilibria because of its relatively large number of coils and their proximity to the plasma. To support its advanced tokamak mission, the DIII–D experimental program continues to push the envelope of this capability, frequently encountering limits imposed by allowable currents in poloidal shaping coils. Violation of current constraints is presently dealt with by operator adjustment of control targets and gains between plasma discharges. At the same time, demands for more precise and stable control have motivated efforts to develop and install advanced multivariable algorithms for control of plasma shape in DIII-D and other devices. Further, there is currently no way to ensure respect of nonlinear current constraints in a multivariable linear controller design and no practical way to manually tune these fully coupled controllers between discharges after installation. Various linear minimization schemes can be implemented to encourage currents to remain within limits, but adherence to these limits cannot be guaranteed by linear methods alone. In this paper, we describe ongoing efforts to provide methods that guarantee currents will not exceed preset limits, and that simultaneously achieve the best obtainable quality of control subject to current limit constraints.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Development of robust and multi-mode control of tearing in DIII-D

Neoclassical tearing modes (NTMs) are instabilities that can produce undesirable magnetic islands in tokamak plasmas. They can be stabilized by applying electron cyclotron current drive (ECCD) at the island. The NTM control system on DIII-D can now control multiple modes. Each of 6 mirrors that reflect ECCD beams into the plasma can be assigned to different surfaces in the plasma where NTMs are unstable. The control system then steers the mirrors to keep the beams aimed at the surfaces. The system routinely stabilizes one NTM preemptively and has now also been used to control two modes in the same discharge. With the “catch-and-subdue” function, ECCD-generating gyrotrons can be turned on when NTMs appear and off after suppression. Newly triggered NTMs can be promptly suppressed if mode onset is detected early and ECCD immediately applied. Early mode detection is achieved in this paper by spectral analysis of Mirnov probes with a band-pass filter for the expected mode frequency. Targeted surfaces are tracked by equilibrium reconstructions (that include measurements of the motional Stark effect). The ECCD position is tracked by ray-tracing using the TORBEAM code. Several techniques are being explored for fine-tuning alignment when NTMs occur. One method adjusts ECCD alignment in steps until the island decays fast enough. A second method sweeps the alignment to find the optimum. A third method pulses gyrotrons and uses electron cyclotron emission to compare where the resulting temperature pulses are relative to temperature fluctuations from a rotating NTM. NTM control in ITER is expected to use active profile regulation to maximize controllability, followed by repeated catch-and-subdue actions if modes are retriggered, in order to maintain island size below the disruptive threshold while maximizing confinement and fusion gain. Between events, real-time tracking will be performed to maintain alignment and readiness for subsequent catch-andsubdue actions. Methods for active probing of stability boundaries will be studied as possible diagnostics for the profile regulation. Finally, selected elements of this ITER NTM control vision will be discussed and assessed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Progress and plan of KSTAR plasma control system upgrade

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. Here, 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗