Progress towards steady state at low aspect ratio on the National Spherical Torus Experiment (NSTX)

Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S equivalent to q95 (I-p/a B-t)) of similar to 41 (MA m- I T- 1) simultaneous with a record plasma elongation of K bla similar to 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance. Plasmas with high shape factor have been sustained for pulse lengths which correspond to tau(pulse),, = 1.6s similar to 50 tau(E) similar to 5 tau(CR), where tau(CR) is the current relaxation time and tau(E) is the energy confinement time. Plasmas with higher similar to 20% have been sustained for tau(pulse), = 1.2s similar to 25TE similar to 3TCR with non-inductive current fractions f(NI) similar to 50%, with similar to 40% pressure driven current and similar to 10% neutral beam driven current. An interesting feature of these discharges is the observation that the central value of the safety factor q (0) remains elevated and constant for several current diffusion times without sawteeth, similar to the 'hybrid mode'. Calculations of the profiles of inductive and non-inductive current are compared with measurements of the total current profile and are shown to be in quantitative agreement. Results are shown from experiments which investigate the applicability of high harmonic fast waves (HHFWs) and electron Bernstein waves (EBWs) as current drive and heating sources, and the possibility of LHCD for future ST devices is raised. A calculated scenario which provides 100% non-inductive current drive is described. NSTX operates with peak divertor heat fluxes which are in the same range as those expected for the ITER device, i.e. with P-heatmax similar to 10MW m(-2). High triangularity, high elongation plasmas on NSTX have been demonstrated to have reduced peak heat flux to the divertor plates to < 3 MW m(-2).