Progress on Transport Studies in the Spherical Tokamak (ST) Pi3 at General Fusion Inc

The ST Pi3 is a medium-size device in operation at General Fusion Inc. aiming to guide the ST Pi4 (currently under construction) and the Fusion Demonstration Plant (FDP) to be constructed at the Culham Laboratory site, UK[1].

The Pi3 plasmas are formed exclusively via coaxial-helicity injection (CHI) and last ~25ms. This plasma can be close-fitted or slightly detached to an Al spherical vacuum vessel of 1.00m inner radius and 50mm thickness, acting as a flux-conserver. The wall is regularly coated with lithium using 4 heated evaporators.

Typical values of Pi3 plasmas in early phases of discharge (~3ms) are: R_o=0.67m, a=0.37m, A=1.8, k=1.7, δ~0.4, B_T(R_o)=0.23T, I_p =300kA, n_e(bar)=4.3x10¹⁹m^-3, T_e(0)=170eV and T_i(bar)=230eV.

Equilibrium reconstruction is performed by Flagship code using constraints from poloidal pickup coils inside of the vessel wall and polarimeter signals[2].

Results of the thermal energy confinement time tauE and the thermal diffusivity Chi will be presented. The former is calculated from kinetics and magnetic power balance methods. Results from kinetics in medium plasma densities (N_G=0.6) show tauE=8ms, close to the expected from the ITER97-L-mode scaling: 9ms. Analysis of new regimes at higher B_T(R_o), I_p, and T_e(0) will also be presented.

[1] A. Mossman et al., in this conference

[2] A. Froese et al, in this conference