FLIPEC, an Ideal MHD free-boundary equilibrium solver for plasma with flows in axisymmetric geometries.

The ability of calculating ideal MHD equilibria is needed not only during the design of a magnetic fusion device, but also in its normal operation. In the case of tokamak axisymmetric configurations with significant plasma flow, the plasma edge is maintained fixed as the code searches for the equilibrium in the majority of codes. However, plasma flows may induce changes such as displacements of the position of the X-point, the magnetic axis or the shape of the plasma boundary, to name a few, that cannot be quantified properly in a fixed-boundary setup. In this contribution we present a new equilibrium code and its most significant and recent upgrades. In this context, FLIPECode was written to obtain free-boundary axisymmetric ideal MHD equilibria for arbitrary plasma flows. Particularly, the free-plasma-boundary scheme has been previously applied to 3D, non-axisymmetric configurations in the absence of flow within the SIESTA equilibrium code.

The upgrades open the code to many other possibilities. FLIPEC was originally written in toroidal coordinates, (r,θ,ξ), providing the code with a circular domain section, what made the code incompatible with many tokamaks, since the coils and solenoid leads to a singularity in the magnetic field when included in the domain. Currently, general coordinates have been implemented and the computational mesh can be adapted to any plasma shape. Additionally, the onset of a vertical instability has been successfully addressed by the use of a control stability scheme which utilizes two stabilization coils.

Examples of application of the code to ITER and NSTX configurations will be used to showcase the capabilities of the code. Additional developments currently in progress, such as the possibility of keeping fixed some geometrical locations (magnetic axis, x-point) and the targeting to significant parameters (integrated current) will be discussed as well.