Development of a new error field correction coil (C-coil) for DIII-D

The C-coil recently installed on the DIII-D tokamak was developed to reduce the error fields created by imperfections in the location and geometry of the existing coils used to confine, heat, and shape the plasma. First results from C-coil experiments include stable operation in a 1.6 MA plasma with a density less than 1.0 {times} 10{sup 13} cm{sup {minus}3}, nearly a factor of three lower density than that achievable without the C-coil. The C-coil has also been used in magnetic braking of the plasma rotation and high energy particle confinement experiments. The C-coil system consists of six individual saddle coils, each 60{degree} wide toroidally, spanning the midplane of the vessel with a vertical height of 1.6 m. The coils are located at a major radius of 3.2 m, just outside of the toroidal field coils. The actual shape and geometry of each coil section varied somewhat from the nominal dimensions due to the large number of obstructions to the desired coil path around the already crowded tokamak. Each coil section consists of four turns of 750 MCM insulated copper cable banded with stainless steel straps within the web of a 3 in. x 3 in. stainless steel angle frame. The C-coil structure was designed to resist peak transient radial forces (up to 1,800 Nm) exerted on the coil by the toroidal and ploidal fields. The coil frames were supported from existing poloidal field coil case brackets, coil studs, and various other structures on the tokamak

Non-linear instability of DIII-D to error fields

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

Real-time protection of the ohmic heating coil force limits in DIII-D

The maximum safe operating limits of the DIII-D tokamak are determined by the force produced in the ohmic heating coil and the toroidal field coil during a plasma pulse. This force is directly proportional to the product of the current in the coils. Historically, the current limits for each coil were set statically before each pulse without regard for the time varying nature of the currents. In order to allow the full time-dependent capability of the ohmic coil to be used, a system was developed for monitoring the product of the currents dynamically and making appropriate adjustments in real time. This paper discusses the purpose, implementation, and results of this work

Investigation of neutral beam arc chamber failure during helium operations at DIII-D

The Neutral Beam system on the DIII-D tokamak consists of eight ion sources using the Common Long Pulse Source (CLPS) design. During helium operation, desired for research regarding the ITER pre-nuclear phase, it has been observed that the ion source arc chamber performance steadily deteriorates, eventually failing due to electrical breakdown across the insulation. This poster presents the details and preliminary results of an experimental effort to replicate the problem in a bench top ion source with similar plasma parameters. The initial aim of the experiment is to test the hypothesis that during helium operation there is increased tungsten evaporation and sputtering due to ion bombardment of the hot cathodes, leading to the deposition of filament material on the insulation and subsequent short circuits. Ultimately the aim of the experiment is to find methods to ameliorate the problems associated with helium operation at DIII-D

Anomalies in the Applied Magnetic Fields on Diii-D and Their Implications for the Understanding of Stability Experiments

Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments and resistive wall mode experiments done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made on DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported. Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all of the coils are one third of the values indicated from the stability experiments. These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks

Experimental studies of the arc chamber short circuit failure mechanism on the DIII-D neutral beam system

Here we report on efforts to improve performance and longevity of the Neutral Beam Injection (NBI) system by initiating a R&D program aimed at studying the most common failure mechanism for the ion sources. To this end a filament driven plasma chamber has been constructed with plasma parameters similar to the arc chamber of NBI ion sources. A preliminary report of an investigation into the most common failure is presented here: The failure mechanism observed during helium operations on DIII-D is the result of electrical breakdown of the insulation material that separates the filament plates from the anode. The fault is reproduced in a table top experiment analogous to the DIII-D NBI ion source in key parameters and proposals for amelioration of the issue are discussed