Abstract
In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ∼ 2.1 to κ ∼ 2.6—approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βt with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented.
Highlights
The spherical torus concept [1] is an extension of the same reasoning that leads to the steadystate advanced tokamak concept
Using the bootstrap current and the external non-inductive current drive to sustain a tokamak in steady state is an immediate precursor to the concept of changing the geometry of the torus to optimize bootstrap current and minimize the need for a transformer to drive current inductively
In the low aspect ratio limit, it becomes impossible to use known superconducting materials to build a toroidal field coil for a reactor due to nuclear heating. This lack of shielding requires low aspect ratio devices to employ conventional conductors, which in turn forces the device to operate at very high toroidal β so as to minimize the recirculating power fraction
Summary
The spherical torus concept [1] (i.e. a tokamak with R/a ∼ 1) is an extension of the same reasoning that leads to the steadystate advanced tokamak concept. The spherical torus concept takes the extreme step of eliminating the transformer and maximizing toroidal field utilization by reducing the physical size of the toroidal field coil to engineering limits. 100% non-inductive current drive at βt ∼ 40% is possible on NSTX NSTX has successfully driven current using the high harmonic fast waves (HHFWs)
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