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Modification of the two existing DIII-D neutral beamlines is planned to allow vertical steering to provide off-axis neutral beam current drive (NBCD) peaked as far off-axis as half the plasma minor radius. New calculations for a downward-steered beam indicate strong current drive with good localization off-axis so long as the toroidal magnetic field, B T , and the plasma current, I p , point in the same direction. This is due to good alignment of neutral beam injection (NBI) with the local pitch of the magnetic field lines. This model has been tested experimentally on DIII-D by injecting equatorially mounted NBs into reduced size plasmas that are vertically displaced with respect to the vessel midplane. The existence of off-axis NBCD is evident in the changes seen in sawtooth behaviour in the internal inductance. By shifting the plasma upwards or downwards, or by changing the sign of the toroidal field, off-axis NBCD profiles measured with motional Stark effect data and internal loop voltage show a difference in amplitude (40-45%) consistent with differences predicted by the changed NBI alignment with respect to the helicity of the magnetic field lines. The effects of NBI direction relative to field line helicity can be large even in ITER: off-axis NBCD can be increased by more than 30% if the B T direction is reversed. Modification of the DIII-D NB system will strongly support scenario development for ITER and future tokamaks as well as provide flexible scientific tools for understanding transport, energetic particles and heating and current drive.

Measurements of magnetomechanical coupling, ΔE effect, and relative permeability have been made as a function of bias field and annealing temperature and time in transversely annealed samples of Metglas R a) alloy 2605S-2 (Fe78Si9B13), Metglas alloy 2605S-3 (Fe79Si5B16), and Metglas alloy 2826MB (Fe40Ni38Mo4B18). Annealing temperatures and times were varied from 300 to 450 °C and 15 to 60 min, respectively. The bias field which generates the largest coupling factor in these alloys is less than 1 Oe. From an application point of view, this low optimum field has a distinct advantage in power requirements over the 11 Oe field required for Metglas alloy 2605CO1 but the low field requirement is a marked disadvantage when considering bias field stability. That is, a 0.5 Oe change in field can result in greatly reduced coupling in the present alloys versus a very small change in 2605CO.

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Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. Recently, these coils have also demonstrated significant benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modeling has shown that substantial performance improvements can be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and have been tested successfully. Installation of a set of twelve internal coils and magnetic sensors in the DIII-D tokamak is to be completed in December 2002.The design requirement for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel.conductor and the stainless tube without overheating the polyamide insulator.The primary challenge in the design of these coils was in joining of both the copper GENERAL ATOMICS REPORT GA-A24056 1 P.M. ANDERSON, et a/.