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Limits on m = 2, n = 1 error field induced locked mode instability in TPX with typical sources of poloidal field coil error field and a prototype correction coil, C-coil''

Description: Irregularities in the winding or alignment of poloidal or toroidal magnetic field coils in tokamaks produce resonant low m, n = 1 static error fields. Otherwise stable discharges can become nonlinearly unstable, and locked modes can occur with subsequent disruption when subjected to modest m = 2, n = 1 external perturbations. Using both theory and the results of error field/locked mode experiments on DIII-D and other tokamaks, the critical m = 2, n = 1 applied error field for locked mode instability in TPX is calculated for discharges with ohmic, neutral beam, or rf heating. Ohmic discharges axe predicted to be most sensitive, but even co-injected neutral beam discharges (at [beta][sub N] = 3) in TPX will require keeping the relative 2, 1 error field (B[sub r21]/B[sub T]) below 2 [times] 10[sup [minus]4]. The error fields resulting from as-built'' alignment irregularities of various poloidal field coils are computed. Coils if well-designed must be positioned to within 3 mm with respect to the toroidal field to keep the total 2,1 error field within limits. Failing this, a set of prototype correction coils is analyzed for use in bringing 2,1 error field down to a tolerable level.
Date: December 1, 1992
Creator: La Haye, R.J.
Partner: UNT Libraries Government Documents Department

Physics of locked modes in ITER: Error field limits, rotation for obviation, and measurement of field errors

Description: The existing theoretical and experimental basis for predicting the levels of resonant static error field at different components m,n that stop plasma rotation and produce a locked mode is reviewed. For ITER ohmic discharges, the slow rotation of the very large plasma is predicted to incur a locked mode (and subsequent disastrous large magnetic islands) at a simultaneous weighted error field ({Sigma}{sub 1}{sup 3}w{sub m1}B{sup 2}{sub rm1}){sup {1/2}}/B{sub T} {ge} 1.9 x 10{sup -5}. Here the weights w{sub m1} are empirically determined from measurements on DIII-D to be w{sub 11} = 0. 2, w{sub 21} = 1.0, and w{sub 31} = 0. 8 and point out the relative importance of different error field components. This could be greatly obviated by application of counter injected neutral beams (which adds fluid flow to the natural ohmic electron drift). The addition of 5 MW of 1 MeV beams at 45{degrees} injection would increase the error field limit by a factor of 5; 13 MW would produce a factor of 10 improvement. Co-injection beams would also be effective but not as much as counter-injection as the co direction opposes the intrinsic rotation while the counter direction adds to it. A means for measuring individual PF and TF coil total axisymmetric field error to less than 1 in 10,000 is described. This would allow alignment of coils to mm accuracy and with correction coils make possible the very low levels of error field needed.
Date: February 1, 1997
Creator: La Haye, R.J.
Partner: UNT Libraries Government Documents Department

Use of an m=2, n=1 static error field correction coil, {open_quotes}The C-coil,{close_quotes} on DIII-D to avoid disruptive locked modes

Description: Minimizing resonant, static n = 1 error field with a phase steerable correction coil the C-coil, in DIII-D allows avoidance of disruptive locked modes. Alternatiely, increasing n = 1 error field in rapidly rotating plasmas can induce magnetic braking of rotation without locking for the study of the role of rotation on stability. Small toroidally asymmetric m = 2, n = 1 static field errors are of concern for the design of next-generation devices and for the operation of existing tokamaks. In low density ohmic plasmas for example, the torque of a small resonant error at the q = 2 surface can overcome the plasma inertial and/or viscous forces, stop the rotation and produce a large island which can cause disruption.
Date: May 1, 1995
Creator: La Haye, R.J. & Scoville, J.T.
Partner: UNT Libraries Government Documents Department

Non-linear instability of DIII-D to error fields

Description: Otherwise stable DIII-D discharges can become nonlinearly unstable to locked modes and disrupt when subjected to resonant m = 2, n = 1 error field caused by irregular poloidal field coils, i.e. intrinsic field errors. Instability is observed in DIII-D when the magnitude of the radial component of the m = 2, n = 1 error field with respect to the toroidal field is B{sub r21}/B{sub T} of about 1.7 {times} 10{sup {minus}4}. The locked modes triggered by an external error field are aligned with the static error field and the plasma fluid rotation ceases as a result of the growth of the mode. The triggered locked modes are the precursors of the subsequent plasma disruption. The use of an n = 1 coil'' to partially cancel intrinsic errors, or to increase them, results in a significantly expanded, or reduced, stable operating parameter space. Precise error field measurements have allowed the design of an improved correction coil for DIII-D, the C-coil'', which could further cancel error fields and help to avoid disruptive locked modes. 6 refs., 4 figs.
Date: October 1, 1991
Creator: La Haye, R.J. & Scoville, J.T.
Partner: UNT Libraries Government Documents Department

Multi-Device Scaling of Neoclassical Tearing Mode Onset with Beta

Description: The islands from tearing modes driven unstable and sustained by helically perturbed neo-classical bootstrap current at high beta often provide the practical limit to long-pulse, high confinement tokamak operation [1,2]. The discharges studied are ELMy H-mode single-null divertor (SND) at q{sub 95} {approx}> 3. Periodic sawteeth with m/n = 1/1 and 2/2 are observed to induce m/n = 3/2 neoclassical tearing modes (NTMs) in the tokamaks Asdex-Upgrade [3], DIII-D [4] and JET [5]; confinement can drop by up to 30%, constituting a ''soft'' beta limit. Data for the onset of these modes was obtained by slowly raising beta on a time scale longer than the sawteeth period and observing the beta value at onset. Comparison of the measured critical beta to a model for the critical beta is made in terms of dimensionless parameters. This modeling is then used for extrapolation/prediction to a reactor-grade tokamak.
Date: July 1, 1999
Creator: La Haye, R.J.; Buttery, R.J.; Guenter, S.; Huysmans, G.T.A. & Wilson, H.R.
Partner: UNT Libraries Government Documents Department

An X-point ergodic divertor

Description: A new ergodic divertor is proposed. It utilizes a system of external (n = 3) coils arranged to generate overlapping magnetic islands in the edge region of a diverted tokamak and connect the randomized field lines to the external (cold) divertor plate. The novel feature in the configuration is the placement of the external coils close to the X-point. A realistic design of the external coil set is studied by using the field line tracing method for a low aspect ratio (A {approx equal} 3) tokamak. Two types of effects are observed. First, by placing the coils close to the X-point, where the poloidal magnetic field is weak and the rational surfaces are closely packed only a moderate amount of current in the external coils is needed to ergodize the edge region. This ergodized edge enhances the edge transport in the X-point region and leads to the potential of edge profile control and the avoidance of edge localized modes (ELMs). Furthermore, the trajectories of the field lines close to the X-point are modified by the external coil set, causing the hit points on the external divertor plates to be randomized and spread out in the major radius direction. A time-dependent modulation of the currents in the external (n = 3) coils can potentially spread the heat flux more uniformly on the divertor plate avoiding high concentration of the heat flux. 10 refs., 9 figs.
Date: October 1, 1991
Creator: Chu, M.S.; Jensen, T.H.; La Haye, R.J.; Taylor, T.S. & Evans, T.E.
Partner: UNT Libraries Government Documents Department

Closed Loop Feedback of MHD Instabilities on DIII-D

Description: A system of coils, sensors and amplifiers has been installed on the DIII-D tokamak to study the physics of feedback stabilization of low-frequency MHD [magnetohydrodynamic] modes such as the Resistive Wall Mode (RWM). Experiments are being performed to assess the effectiveness of this minimal system and benchmark the predictions of theoretical models and codes. In the last campaign, the experiments have been extended to a regime where the RWM threshold is lowered by a fast ramp of the plasma current. In these experiments, the onset time of the RWM is very reproducible. With this system, the onset of the RWM has been delayed by up to 100 msec without degrading plasma performance. The growth rate of the mode increases proportional to the length of delay, suggesting that the plasma is evolving towards a more unstable configuration. The present results have suggested directions for improving the feedback system including better sensors and improved feedback algorithms.
Date: January 16, 2001
Creator: Fredrickson, E.D.; Bialek, J.; Garofalo, A.M.; Johnson, L.C.; La Haye, R.J. & Lazarus, E.A.
Partner: UNT Libraries Government Documents Department

Relationship Between Onset Thresholds, Trigger Types, and Rotation Shear for the m/n=2/1 Neoclassical Tearing Mode in a High-β Spherical Torus

Description: The onset conditions for the m/n=2/1 neoclassical tearing mode (NTM) are studied in terms of neoclassical drive, triggering instabilities, and toroidal rotation or rotation shear, in the spherical torus NSTX [M. Ono, et al., Nuclear Fusion 40, 557 (2000)]. There are three typical onset conditions for these modes, given in order of increasing neoclassical drive required for mode onset: triggering by energetic particle modes, triggering by edge localized modes, and cases where the modes appear to grow without a trigger. In all cases, the required drive increases with toroidal rotation shear, implying a stabilizing effect from the shear.
Date: February 24, 2009
Creator: Gerhardt, S. P.; Brennan, D. P.; Buttery, R.; La Haye, R. J.; Sabbagh, S.; Strait, E. et al.
Partner: UNT Libraries Government Documents Department

Neoclassical tearing modes in DIII-D and calculations of the stabilizing effects of localized electron cyclotron current drive

Description: Neoclassical tearing modes are found to limit the achievable beta in many high performance discharges in DIII-D. Electron cyclotron current drive within the magnetic islands formed as the tearing mode grows has been proposed as a means of stabilizing these modes or reducing their amplitude, thereby increasing the beta limit by a factor around 1.5. Some experimental success has been obtained previously on Asdex-U. Here the authors examine the parameter range in DIII-C in which this effect can best be studied.
Date: May 1, 1999
Creator: Prater, R.; La Haye, R.J.; Lin-Liu, Y.R.; Lohr, J.; Bernabei, S.; Perkins, F.W. et al.
Partner: UNT Libraries Government Documents Department

Practical beta limit in ITER-shaped discharges in DIII-D and its increase by higher collisionality

Description: The maximum beta which can be sustained for a long pulse in ITER-shaped plasmas in DIII-D with q{sub 95} {approx_gt} 3, ELMs, and sawteeth is found to be limited by resistive tearing modes, particularly m/n = 3/2 and 2/1. At low collisionality comparable to that which will occur in ITER, the beta limit is a factor of two below the usually expected n = {infinity} ballooning and n = 1 kink ideal limits.
Date: October 1, 1996
Creator: La Haye, R.J.; Chu, M.S. & Callen, J.D.
Partner: UNT Libraries Government Documents Department

Wall stabilization of high beta plasmas in DIII-D

Description: Detailed analysis of recent high beta discharges in the DIII-D tokamak demonstrates that the resistive vacuum vessel can provide stabilization of low n magnetohydrodynamic (MHD) modes. The experimental beta values reaching up to {beta}{sub T} = 12.6% are more than 30% larger than the maximum stable beta calculated with no wall stabilization. Plasma rotation is essential for stabilization. When the plasma rotation slows sufficiently, unstable modes with the characteristics of the predicted {open_quotes}resistive wall{close_quotes} mode are observed. Through slowing of the plasma rotation between the q = 2 and q = 3 surfaces with the application of a non-axisymmetric field, the authors have determined that the rotation at the outer rational surfaces is most important, and that the critical rotation frequency is of the order of {Omega}/2{pi} = 1 kHz.
Date: February 1, 1995
Creator: Taylor, T.S.; Strait, E.J.; Lao, L.L.; Turnbull, A.D.; Burrell, K.H.; Chu, M.S. et al.
Partner: UNT Libraries Government Documents Department

SUSTAINED STABILIZATION OF THE RESISTIVE WALL MODE BY PLASMA ROTATION IN THE DIII-D TOKAMAK

Description: OAK-B135 A path to sustained stable operation, at plasma pressure up to twice the ideal magnetohydrodynamic (MHD) n = 1 free-boundary pressure limit, has been discovered in the DIII-D tokamak. Tuning the correction of the intrinsic magnetic field asymmetries so as to minimize plasma rotation decay during the high beta phase and increasing the angular momentum injection, have allowed maintaining the plasma rotation above that needed for stabilization of the resistive wall mode (RWM). A new method to determine the improved magnetic field correction uses feedback to sense and minimize the resonant plasma response to the non-axisymmetric field. At twice the free-boundary pressure limit, a disruption precursor is observed, which is consistent with having reached the ''ideal wall'' pressure limit predicted by stability calculations.
Date: October 1, 2001
Creator: GAROFALO,A.M; STRAIT,E.J; JOHNSON,L.C; LA HAYE,R.J; LAZARUS,E.A; NAVRATIL,G.A et al.
Partner: UNT Libraries Government Documents Department

STABILZATION OF NEOCLASSICAL TEARING MODES BY LOCALIZED ECCD IN DIII-D

Description: Neoclassical tearing modes (NTMs) are MHD modes which can become destabilized in a tokamak by a helical pressure perturbation. 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 and the 2/1 tearing modes 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 flux surface of maximum mode amplitude. When the mode first appears in the plasma the stored energy decreases by 30%, but when 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 effecting 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. Only partial success has been obtained in stabilizing the 2/1 mode by ECCD, but calculations indicate that ECCD power near 3 MW should be adequate for complete suppression of this mode.
Date: May 1, 2002
Creator: Prater, R.; Eliss, R. A., III; La Haye, R. J.; Lohr, J. M.; Luce, T. C.; Perkins, F. W. et al.
Partner: UNT Libraries Government Documents Department

Optimization of negative central shear discharges in shaped cross sections

Description: Magnetohydrodynamic (MHD) stability analyses of Negative Central Shear (NCS) equilibria have revealed a new understanding of the limiting MHD instabilities in NCS experiments. Ideal stability calculations show a synergistic effect between cross section shape and pressure profile optimization; strong shaping and broader pressure independently lead to moderately higher {Beta} limits, but broadening of the pressure profile in a strongly dee-shaped cross- section leads to a dramatic increase in the ideal {Beta} limit. Localized resistive interchange (RI) modes can be unstable in the negative shear region and are most restrictive for peaked pressure profiles. Resistive global modes can also be destabilized significantly below the ideal P limit. Experiments largely confirm the general trends, and diagnostic measurements and numerical stability calculations are found to be in good qualitative agreement. Observed disruptions in NCS discharges with L-mode edge and strongly peaked pressure, appear to be initiated by interactions between the RI, and the global ideal and resistive modes.
Date: October 1, 1996
Creator: Turnbull, A.D., Chu, M.S., Taylor, T.S., Casper, T.A., Rice, B.W.; Greene, J.M., Greenfield, C.M., La Haye, R.J., Lao, L.L., Lee, B.J.; Miller, R.L., Ren, C., Strait, E.J., Tritz, K.; Rettig, C.L., Rhodes, T.L. & Sauter, O.
Partner: UNT Libraries Government Documents Department

A Comparison of Plasma Performance Between Single-Null and Double-Null Configurations During Elming H-Mode

Description: Tokamak plasma performance generally improves with increased shaping of the plasma cross section, such as higher elongation and higher triangularity. The stronger shaping, especially higher triangularity, leads to changes in the magnetic topology of the divertor. Because there are engineering and divertor physics issues associated with changes in the details of the divertor flux geometry, especially as the configuration transitions from a single-null (SN) divertor to a marginally balanced double-null (DN) divertor, we have undertaken a systematic evaluation of the plasma characteristics as the magnetic geometry is varied, particularly with respect to (1) energy confinement, (2) the response of the plasma to deuterium gas fueling, (3) the operational density range for the ELMing H-mode, and (4) heat flux sharing by the diverters. To quantify the degree of divertor imbalance (or equivalently, to what degree the shape is double-null or single-null), we define a parameter DRSEP. DRSEP is taken as the radial distance between the upper divertor separatrix and the lower divertor separatrix, as determined at the outboard midplane. For example, if DRSEP=O, the configuration is a magnetically balanced DN; if DRSEP = +1.0 cm, the divertor configuration is biased toward the upper divertor. Three examples are shown in Fig. 1. In the following discussions, VB drift is directed toward the lower divertor.
Date: July 1, 1999
Creator: Petrie, T.W.; Fenstermacher, M.E.; Allen, S.L.; Carlstrom, T.N.; Gohil, P.; Groebner, R.J. et al.
Partner: UNT Libraries Government Documents Department

ELM Suppression in Low Edge Collisionality H-Mode Discharges Using n=3 Magnetic Perturbations

Description: Using resonant magnetic perturbations with toroidal mode number n = 3, we have produced H-mode discharges without edge localized modes (ELMs) which run with constant density and radiated power for periods up to about 2550 ms (17 energy confinement times). These ELM suppression results are achieved at pedestal collisionalities close to those desired for next step burning plasma experiments such as ITER and provide a means of eliminating the rapid erosion of divertor components in such machines which could be caused by giant ELMs. The ELM suppression is due to an enhancement in the edge particle transport which reduces the edge pressure gradient and pedestal current density below the threshold for peeling-ballooning modes. These n = 3 magnetic perturbations provide a means of active control of edge plasma transport.
Date: July 11, 2005
Creator: Burrell, K H; Evans, T E; Doyle, E J; Fenstermacher, M E; Groebner, R J; Leonard, A W et al.
Partner: UNT Libraries Government Documents Department