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OBSERVATION OF SELF-MITIGATION OF A DENSITY LIMIT DISRUPTION IN DIII-D

Description: OAK-B135 Density limit disruptions set an upper bound on the electron density in tokamaks and are important for future reactor-size tokamaks, which will typically need to operate at high densities to achieve ignition. In the standard picture of disruptions, a large MHD mode, or combination of MHD modes, causes a mixing of previously nested magnetic flux surfaces across much of the profile. Rapid heat and particle transport across the separatrix result, and the thermal energy of the discharge is lost along open field lines into the divertor on a millisecond time scale or faster. In this work, a density limit disruption is initiated by ramping up the density in a lower single-null discharge in the DIII-D tokamak. As in most disruptions, a large MHD precursor is observed. However, in contrast with the disruption scenario described above, it is found that the plasma thermal energy, rather than being conducted into the divertor, is dominantly lost by radiation to the main chamber walls. This has been referred to as self-mitigation of the disruption, in comparison to the intentional mitigation of localized heat loads in disruptions by the introduction of pellets or liquid or gas jets to enhance radiation. The self-mitigation effect appears to result from a release of neutrals (deuterium and carbon) from the graphite vacuum vessel walls. These results could have favorable implications for the severity of divertor heat loads during density limit disruptions in future large tokamaks.
Date: August 1, 2003
Creator: GRAY,DS; HOLLMANN,EM; WHYTE,DG; PIGAROV,AYu; KRASHENINNIKOV,SI; BOEDO,JA et al.
Partner: UNT Libraries Government Documents Department

Multivariable shape control development on the DIII-D tokamak

Description: In this paper, the authors describe recent work on plasma shape and position control at DIII-D. This control consists of two equally challenging problems--the problem of identifying what the plasma actually looks like in real time, i.e. measuring the parameters to be controlled, and the task of determining the feedback algorithm which best controls these plasma parameters in a multiple input-output system. Recent implementation of the EFIT plasma equilibrium reconstruction algorithm in real time code which produces a new equilibrium estimate every 1.5 ms seems to solve the longstanding problem of obtaining sufficiently accurate plasma shape and position estimation. Stabilization of the open-loop unstable vertical motion is also viewed as a solved problem. The primary remaining problem appears to be how best to command the power supplies to achieve a desired shaping control response. They will describe the effort to understand and apply linearized models of plasma evolution to development and implementation of multivariable plasma controllers.
Date: November 1, 1997
Creator: Walker, M.L.; Humphreys, D.A. & Ferron, J.R.
Partner: UNT Libraries Government Documents Department

Control of plasma poloidal shape and position in the DIII-D tokamak

Description: Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks.
Date: November 1, 1997
Creator: Walker, M.L.; Humphreys, D.A. & Ferron, J.R.
Partner: UNT Libraries Government Documents Department

NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK

Description: OAK A271 NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK. The advanced tokamak (AT) operating mode which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. Simultaneous high performance regulation of the plasma boundary and internal profiles requires multivariable control techniques to account for the highly coupled influences of equilibrium shape, profile, and stability control. This paper describes progress towards the DIII-D At mission goal through both significantly improved real-time computational hardware and control algorithm capability.
Date: October 1, 2002
Creator: WALKER, ML; FERRON, JR; HUMPHREYS, DA; JOHNSON, RD; LEUER, JA; PENAFLOR, BG et al.
Partner: UNT Libraries Government Documents Department

INCREASED STABLE BETA IN DIII-D BY SUPPRESSION OF A NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE AND ACTIVE FEEDBACK

Description: OAK A271 INCREASED STABLE BETA IN DIII-D BY SUPPRESSION OF A NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE AND ACTIVE FEEDBACK. In DIII-D, the first real-time active control of the electron cyclotron current drive stabilization of a neoclassical tearing mode (here m/n=3/2) is demonstrated. The plasma control system is put into a search and suppress mode to align the ECCD with the island by making either small rigid radial position shifts (of order 1 cm) of the entire plasma (and thus the island) or small changes in toroidal field (of order 0.5%) which radially moves the second harmonic resonance location (and thus the rf current drive). The optimum position minimizes the real-time mode amplitude signal and stabilization occurs despite changes in island location from discharge-to-discharge or from time-to-time. When the neutral beam heating power is programmed to rise after mode suppression by the ECCD, the plasma pressure increases above the peak at the onset of the neoclassical tearing mode until the magnetic island reappears due to the ECCD no longer being on the optimal position. Real-time tracking of the change in location of q=3/2 due to the Shafranov shift with increasing beta is necessary to position the ECCD in the absence of a mode so that higher stable beta can be sustained. The control techniques developed for the m/n=3/2 NTM are also being applied to the more deleterious m/n-2/1 NTM. For the first time in any tokamak, an m/n=2/1 mode has been completely suppressed using radially localized off-axis ECCD.
Date: September 1, 2002
Creator: LAHAYE,RJ; HUMPHREYS,DA; LOHR,J; LUCE,TC; PERKINS,FW; PETTY,CC et al.
Partner: UNT Libraries Government Documents Department

COMPLETE SUPPRESSION OF THE M=2/N-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D

Description: A271 COMPLETE SUPPRESSION OF THE M=2/N-1 NEOCLASSICAL TEARING MODE USING ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D. The first suppression of the important and deleterious m=2/n=1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace the ''missing'' bootstrap current in the island O-point. Experiments on the DIII-D tokamak verify the maximum shrinkage of the m=2/n=1 island occurs when the ECCD location coincides with the q = 2 surface. The DIII-D plasma control system is put into search and suppress mode to make small changes in the toroidal field to find and lock onto the optimum position, based on real time measurements of dB{sub {theta}}/dt, for complete m=2/n=1 NTM suppression by ECCD. The requirements on the ECCD for complete island suppression are well modeled by the modified Rutherford equation for the DIII-D plasma conditions.
Date: March 1, 2003
Creator: PETTY,CC; LAHAYE,LA; LUCE,TC; HUMPHREYS,DA; HYATT,AW; PRATER,R et al.
Partner: UNT Libraries Government Documents Department

COMPLETE SUPPRESSION OF THE M/N = 2/1 NEOCLASSICAL TEARING MODE USING RADIALLY LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D AND THE REQUIREMENTS FOR ITER

Description: A271 COMPLETE SUPPRESSION OF THE M/N = 2/1 NEOCLASSICAL TEARING MODE USING RADIALLY LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D AND THE REQUIREMENTS FOR ITER. DIII-D experiments demonstrate the first real-time feedback control of the relative location of a narrow beam of microwaves to completely suppress and eliminate a growing tearing mode at the q = 2 surface. long wavelength tearing modes such as the m/n = 2/1 instability are particularly deleterious to tokamak operation. Confinement is seriously degraded by the island, plasma rotation can cease (mode-lock) and disruption can occur. The neoclassical tearing mode (NTM) becomes unstable due to the presence of a helically-perturbed bootstrap current and can be stabilized by replacing the missing bootstrap current in the island O-point by precisely located co-electron cyclotron current drive (ECCD). The optimum position is found when the DIII-D plasma control system (PCS) is put into a search and suppress mode that makes small radial shifts (in about 1 cm steps) in the ECCD location based on minimizing the Mirnov amplitude. Requirements for ITER are addressed.
Date: July 1, 2003
Creator: LAHAYE,RJ; LUCE,TC; PETTY,CC; HUMPHREYS,DA; HYATT,AW; PERKINS,FW et al.
Partner: UNT Libraries Government Documents Department

Magnetic and thermal energy flow during disruptions in DIII-D

Description: The authors present results from disruption experiments where they measure magnetic energy flow across a closed surface surrounding the plasma using a Poynting flux analysis to measure the electromagnetic power, bolometers to measure radiation power and IR scanners to measure radiation and particle heat conduction to the divertor. The initial and final stored energies within the volume are found using the full equilibrium reconstruction code EFIT. From this analysis they calculate an energy balance and find that they can account for all energy deposited on the first wall and the divertor to within about 10%.
Date: July 1, 1996
Creator: Hyatt, A.W.; Lee, R.L.; Humphreys, D.A.; Kellman, A.G.; Taylor, P.L.; Cuthbertson, J.W. et al.
Partner: UNT Libraries Government Documents Department

Development of a Closed Loop Simulator for Poloidal Field Control in DIII-D

Description: The design of a model-based simulator of the DIII-D poloidal field system is presented. The simulator is automatically configured to match a particular DIII-D discharge circuit. The simulator can be run in a data input mode, in which prior acquired DIII-D shot data is input to the simulator, or in a stand-alone predictive mode, in which the model operates in closed loop with the plasma control system. The simulator is used to design and validate a multi-input-multi-output controller which has been implemented on DIII-D to control plasma shape. Preliminary experimental controller results are presented.
Date: November 1, 1999
Creator: Leuer, J.A.; Walker, M.L.; Humphreys, D.A.; Ferron, J.R.; Nerem, A. & Penaflor, B.G.
Partner: UNT Libraries Government Documents Department

DISCHARGE IMPROVEMENT THROUGH CONTROL OF NEOCLASSICAL TEARING MODES BY LOCALIZED ECCD IN DIII-D

Description: A271 DISCHARGE IMPROVEMENT THROUGH CONTROL OF NEOCLASSICAL TEARING MODES BY LOCALIZED ECCD IN DIII-D. Neoclassical tearing modes (NTMs) are MHD modes which can limit the performance of high beta discharges in tokamaks, in some cases leading to a major disruption. The destabilizing effect which results in NTM growth is a helical decrease in the bootstrap current caused by a local reduction of the plasma pressure gradient by seed magnetic islands. The NTM is particularly well suited to control since the mode is linearly stable although nonlinearly unstable, so if the island amplitude can be decreased below a threshold size the mode will decay and vanish. One means of shrinking the island is the replacement of the missing bootstrap current by a localized current generated by electron cyclotron current drive (ECCD). This method has been applied to the m=3/n=2 neoclassical tearing mode in DIII-D, in H-mode plasmas with ongoing ELMs and sawteeth, both of which generate seed islands periodically. In the case of the 3/2 mode, full suppression was obtained robustly by applying about 1.5 MW of ECCD very near the rational surface of the mode. When the mode first appears in the plasma the stored energy decreases by 20%, but after the mode is stabilized by the ECCD the beta may be raised above the initial threshold pressure by 20% by additional neutral beam heating, thereby generating an improvement in the limiting beta of nearly a factor 2. An innovative automated search algorithm was implemented to find and retain the optimum location for the ECCD in the presence of the mode.
Date: October 1, 2002
Creator: PRATER,R; LAHAYE,RJ; LOHR,J; LUCE,TC; PETTY,CC; FERRON,JR et al.
Partner: UNT Libraries Government Documents Department

DISRUPTION MITIGATION WITH HIGH-PRESSURE NOBLE GAS INJECTION

Description: OAK A271 DISRUPTION MITIGATION WITH HIGH-PRESSURE NOBLE GAS INJECTION. High-pressure gas jets of neon and argon are used to mitigate the three principal damaging effects of tokamak disruptions: thermal loading of the divertor surfaces, vessel stress from poloidal halo currents and the buildup and loss of relativistic electrons to the wall. The gas jet penetrates as a neutral species through to the central plasma at its sonic velocity. The injected gas atoms increase up to 500 times the total electron inventory in the plasma volume, resulting in a relatively benign radiative dissipation of >95% of the plasma stored energy. The rapid cooling and the slow movement of the plasma to the wall reduce poloidal halo currents during the current decay. The thermally collapsed plasma is very cold ({approx} 1-2 eV) and the impurity charge distribution can include > 50% fraction neutral species. If a sufficient quantity of gas is injected, the neutrals inhibit runaway electrons. A physical model of radiative cooling is developed and validated against DIII-D experiments. The model shows that gas jet mitigation, including runaway suppression, extrapolates favorably to burning plasmas where disruption damage will be more severe. Initial results of real-time disruption detection triggering gas jet injection for mitigation are shown.
Date: October 1, 2002
Creator: WHYTE, DG; JERNIGAN, TC; HUMPHREYS, DA; HYATT, AW; LASNIER, CJ; PARKS, PB et al.
Partner: UNT Libraries Government Documents Department

COMPLETE SUPPRESSION OF THE m/n=2/1 NEOCLASSICAL TEARING MODE USING RADIALLY LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE ON DIII-D AND THE REQUIREMENTS FOR ITER

Description: OAK-B135 DIII-D experiments demonstrate the first real-time feedback control of the relative location of a narrow beam of microwaves to completely suppress and eliminate a growing tearing mode at the q=2 surface. long wavelength tearing modes such as the m/n = 2/1 instability are particularly deleterious to tokamak operation. Confinement is seriously degraded by the island, plasma rotation can cease (mode-lock) and disruption can occur. The neoclassical tearing mode (NTM) becomes unstable due to the presence of a helically-perturbed bootstrap current and can be stabilized by replacing the missing bootstrap current in the island O-point by precisely located co-electron cyclotron current drive (ECCD). The geometry for the ECCD launch, the second harmonic resonance 2f{sub ce} and the q=2 surface are shown. The optimum position is found when the DIII-D plasma control system (PCS) is put into a search and suppress mode that makes small radial shifts (in about 1 cm steps) in the ECCD location based on minimizing the Mirnov amplitude.
Date: June 1, 2003
Creator: LA HAYE,RJ; LUCE,TC; PETTY,CC; HUMPHREYS,DA; HYATT,AW; PERKINS,FW et al.
Partner: UNT Libraries Government Documents Department

PHYSICS AND CONTROL OF ELMING H-MODE NEGATIVE CENTRAL SHEAR ADVANCED TOKAMAK SCENARIO BASED ON EXPERIMENTAL PROFILES FOR ITER

Description: A271 PHYSICS AND CONTROL OF ELMING H-MODE NEGATIVE CENTRAL SHEAR ADVANCED TOKAMAK SCENARIO BASED ON EXPERIMENTAL PROFILES FOR ITER. Key DIII-D AT experimental and modeling results are applied to examine the physics and control issues for ITER to operate in a negative central shear (NCS) AT scenario. The effects of a finite edge pressure pedestal and current density are included based on the DIII-D experimental profiles. Ideal and resistive stability analyses indicate that feedback control of resistive wall modes by rotational drive or flux conserving intelligent coils is crucial for these AT configurations to operate at attractive {beta}{sub N} values in the range of 3.0-3.5. Vertical stability and halo current analyses show that reliable disruption mitigation is essential and mitigation control using an impurity gas can significantly reduce the local mechanical stress to an acceptable level. Core transport and turbulence analyses demonstrate that control of the rotational shear profile is essential to maintain the good confinement necessary for high {beta}. Consideration of edge stability and core transport suggests that a sufficiently wide pedestal is necessary for the projected fusion performance. Heat flux analyses indicate that with core-only radiation enhancement the outboard peak divertor heat load is near the design limit of 10 MW/m{sup 2}
Date: November 1, 2002
Creator: LAO,LL; CHAN,VS; EVANS,TE; HUMPHREYS,DA; LEUER,JA; MAHDAVI,MA et al.
Partner: UNT Libraries Government Documents Department

PHYSICS PROCESSES IN DISRUPTION MITIGATION USING MASSIVE NOBLE GAS INJECTION

Description: Methods for detecting imminent disruptions and mitigating disruption effects using massive injection of noble gases (He, Ne, or Ar) have been demonstrated on the DIII-D tokamak [1]. A jet of high injected gas density (> 10{sup 24} m{sup -3}) and pressure (> 20 kPa) penetrates the target plasma at the gas sound speed ({approx}300-500 m/s) and increases the atom/ion content of the plasma by a factor of > 50 in several milliseconds. UV line radiation from the impurity species distributes the plasma energy uniformly on the first wall, reducing the thermal load to the divertor by a factor of 10. Runaway electrons are almost completely eliminated by the large density of free and bound electrons supplied by the gas injection. The small vertical plasma displacement before current quench and high ratio of current decay rate to vertical growth rate result in a 75% reduction in peak halo current amplitude and attendant forces.
Date: July 1, 2002
Creator: HUMPHREYS, D.A.; WHYTE, D.G.; JERNIGAN, T.C.; T.E.EVANS; GRAY, D.S.; HOLLMANN, E.M. et al.
Partner: UNT Libraries Government Documents Department

Real time equilibrium reconstruction for control of the discharge in the DIII-D tokamak

Description: Optimum performance of a tokamak discharge requires accurate feedback control of many of the discharge parameters. For this to be possible, the values of these parameters must be accurately measured. The values of many discharge parameters, such as shape and safety factor profile, are not directly measured but can be evaluated from the available diagnostic data: magnetic field and flux measurements, for example. The most complete evaluation comes from a least squares fit of the diagnostic data to the Grad-Shafranov model that describes the force balance of the tokamak equilibrium, while allowing for a distributed current source. This full reconstruction of the equilibrium has normally been performed offline using a computation-intensive fitting code such as EFIT. This paper provides an introduction to a practical method for performing an equilibrium reconstruction in real time for arbitrary time-varying discharge shapes and current profiles. A detailed description of the algorithm is given in Ref. 2. An approximate solution to the Grad-Shafranov equilibrium relation is found which best fits the diagnostic measurements so that an equilibrium solution consistent with force balance, expressed in terms of the spatial distributions of the toroidal current density and poloidal flux, is available in real time for accurate evaluation of the discharge parameters. The algorithm is very close to that of EFIT and is executed on a time scale fast enough for control of the DIII-D tokamak.
Date: July 1, 1997
Creator: Ferron, J.R.; Walker, M.L.; Lao, L.L.; Penaflor, B.G.; St. John, H.E.; Humphreys, D.A. et al.
Partner: UNT Libraries Government Documents Department

CONTROL OF NEOCLASSICAL TEARING MODES IN DIII-D

Description: The development of techniques for neoclassical tearing mode (NTM) suppression or avoidance is crucial for successful high beta/high confinement tokamaks. Neoclassical tearing modes are islands destabilized and maintained by a helically perturbed bootstrap current and represent a significant limit to performance at higher poloidal beta. The confinement-degrading islands can be reduced or completely suppressed by precisely replacing the ''missing'' bootstrap current in the island O-point or by interfering with the fundamental helical harmonic of the pressure. Implementation of such techniques is being studied in the DIII-D tokamak [J.L. Luxon, et al., Plasma Phys. and Control. Fusion Research, Vol. 1 (International Atomic Energy Agency, Vienna, 1987) p. 159] in the presence of periodic q = 1 sawtooth instabilities, a reactor relevant regime. Radially localized off-axis electron cyclotron current drive (ECCD) must be precisely located on the island. In DIII-D the plasma control system is put into a ''search and suppress'' mode to make either small rigid radial position shifts of the entire plasma (and thus the island) or small changes in toroidal field (and thus, ECCD location) to find and lock onto the optimum position for complete island suppression by ECCD. This is based on real-time measurements of an m/n = 3/2 mode amplitude dB{sub {theta}}/dt. The experiment represents the first use of active feedback control to provide continuous, precise positioning. An alternative to ECCD makes use of the six toroidal section ''C-Coil'' on DIII-D to provide a large non-resonant static m = 1, n = 3 helical field to interfere with the fundamental harmonic of an m/n = 3/2 NTM. While experiments show success in inhibiting the NTM if a large enough n = 3 field is applied before the island onset, there is a considerable plasma rotation decrease due to n = 3 ''ripple''.
Date: November 1, 2001
Creator: HAYE, R.J. LA; GUNTER, S.; HUMPHREYS, D.A.; LOHR, J.; LUCE, T.C.; MARASCHEK, M.E. et al.
Partner: UNT Libraries Government Documents Department

ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

Description: A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response.
Date: October 1, 2002
Creator: HUMPHREYS,DA; FERRON,JR; GAROFALO,AM; HYATT,AW; JERNIGAN,TC; JOHNSON,RD et al.
Partner: UNT Libraries Government Documents Department

CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES

Description: OAK-B135 Advanced tokamak research in DIII-D seeks to optimize the tokamak approach for fusion energy production, leading to a compact, steady state power source. High power density implies operation at high toroidal beta, {beta}{sub T}=<p>2{micro}{sub 0}/B{sub T}{sup 2}, since fusion power density increases roughly as the square of the plasma pressure. Steady-state operation with low recirculating power for current drive implies operation at high poloidal beta, {beta}{sub P} = <p>2{micro}{sub 0}/<B{sub P}>{sup 2}, in order to maximize the fraction of self-generated bootstrap current. Together, these lead to a requirement of operation at high normalized beta, {beta}{sub N} = {beta}{sub T}(aB/I), since {beta}{sub P}{beta}{sub T} {approx} 25[(1+{kappa}{sup 2})/2] ({beta}{sub N}/100){sup 2}. Plasmas with high normalized beta are likely to operate near one or more stability limits, so control of MHD stability in such plasmas is crucial.
Date: June 1, 2003
Creator: STRAIT,EJ; BIALEK,J; CHANCE,MS; CHU,MS; EDGELL,DH; FERRON,JR et al.
Partner: UNT Libraries Government Documents Department