Francesco A. Volpe

Although modern fusion devices consider the use of lithium as a plasma-facing component, its transport mechanism remains under-investigated, being different from those of heavy impurities such as tungsten. For lithium and other light impurities, turbulent transport dominates over neoclassical contributions, motivating a dedicated study on the determination of turbulent diffusion, thermodiffusion, and convective pinch velocity. In this work, we benchmark a new method to extract these coefficients using the global full-f gyrokinetic GYSELA code. In the absence of a transport barrier, the method reproduces results consistent with helium, showing that thermodiffusion is comparable in magnitude to both diffusion and pure convection. Introducing an E×B shear layer to form a transport barrier modifies the radial structure of the transport coefficients. Transport levels decrease in the inner core, while within the barrier, convection becomes dominant, and all fluxes undergo sign reversals. Diffusive and thermodiffusive contributions are also found to partially compensate each other depending on local gradients. Finally, an evaluation of the peaking factor further demonstrates that the presence of a transport barrier significantly mitigates core impurity accumulation.

Compact nuclear fusion reactors will require advanced plasma-facing components capable of withstanding the higher heat and neutron fluxes characteristic of these machines. Liquid metal walls, based for instance on lithium, have emerged as a promising alternative to solid walls. This study investigates lithium redeposition on plasma-facing walls in a deuterium plasma environment using a self-consistent 1D3V Particle-in-Cell code. The model accounts for the interplay between lithium evaporation and the electric potential with a simplified but relevant physics. Four key parameters were analyzed for their impact on lithium redeposition: the ratio γs of the Li+ outflux over the D+ influx, the magnetic field inclination θ, the ionization mean free path Λiz, and the magnetic field intensity B. Results indicate that γs and θ significantly influence redeposition, leading to redeposition rates higher than 95%, while variations in Λiz and B had secondary effects. In general, an increase in one of these parameters drives to a lower redeposition, except for θ, where potential “bumps” formed under specific conditions, alters ion trajectories and results in a non-monotonic redeposition dependency. The role of lithium in flattening the potential profile and its evaporation rate highlight the need to take Li impurity into account in the calculation of the potential profile.

High temperature superconductors (HTSs) expand the design space of stellarator power-plants (PPs) toward high magnetic fields B, enabling compact major radii R. The present paper scans the space of B, R and other design parameters, finding solutions that are promising from a physics and engineering standpoint, while minimizing the capital cost of the PP and the levelized cost of fusion electricity. Similarly, it identifies minimum-cost design points for next-step burning plasma stellarator experiments of fusion gain 1 < Q < 10 . The study assumes advanced stellarator configurations of reduced aspect ratio, heated by neutral beam injection. Plasma-facing, flowing liquid metal (LM) walls protect it from high heat and neutron fluxes. The study relies on analytical first-principle calculations, and established zero-dimensional (0D) empirical scaling laws. Power flows are illustrated by Sankey diagrams. Plasma operating contours are used to determine the reactor’s start-up path. Sensitivity analyses are conducted to identify the most critical reactor parameters within physics, engineering and costing, quantifying their influence on the economics of the PPs. Such 0D study suggests that the assumed next generation HTS, flowing LM walls, and advances in compact plasma configurations could lead to an ignited stellarator PP of aspect ratio A ∼ 4 , R ⩽ 4 m, B > 9 T, and normalized plasma pressure β ∼ 5 % which would minimize both the cost of electricity and capital cost while achieving a net electric power of about 1 GW.

Magnetic confinement devices for nuclear fusion can be large and expensive. Compact stellarators are promising candidates for cost-reduction, but introduce new difficulties: confinement in smaller volumes requires higher magnetic field, which calls for higher coil-currents and ultimately causes higher Laplace forces on the coils—if everything else remains the same. This motivates the inclusion of force reduction in stellarator coil optimization. In the present paper we consider a coil winding surface, we prove that there is a natural and rigorous way to define the Laplace force (despite the magnetic field discontinuity across the current-sheet), we provide examples of cost associated (peak force, surface-integral of the force squared) and discuss easy generalizations to parallel and normal force-components, as these will be subject to different engineering constraints. Such costs can then be easily added to the figure of merit in any multi-objective stellarator coil optimization code. We demonstrate this for a generalization of the REGCOIL code (Landreman 2017 Nucl. Fusion 57 046003), which we rewrote in python, and provide numerical examples for the NCSX (Zarnstorff et al 2001 Plasma Phys. Control. Fusion 43 A237–49) (now QUASAR) design. We present results for various definitions of the cost function, including peak force reductions by up to 40%, and outline future work for further reduction.

Helical equilibria can be generated by arrangements of planar coils similar to tokamaks, but without a central solenoid and with the toroidal field (TF) coils tilted with respect to the vertical. This is known from earlier numerical works, e.g. Moroz 1995 Phys. Plasmas 2 4269. However, such concept tends to need large coils but form small plasmas (of large aspect ratio). Here it is numerically shown that larger, more attractive vacuum flux surfaces—relative to the size of the device—can be generated by carefully optimizing the inclination of the TF coils and currents in the various coil-sets. Vacuum configurations of aspect ratios as low as 4.1 are found for six tilted TF circular coils. Higher numbers of TF coils have advantages (smaller effective ripple) and disadvantages (lower rotational transform, smaller plasma). Finally, the aspect ratio A of the vacuum flux surfaces is quantified as a function of the ratio A _c of the coil radius to the radial location of the coil center. It is found that, in order to minimize A , it is beneficial to interlink or marginally interlink the TF coils ( A _c ≲ 1).

It is shown that static error fields (EFs) can severely limit the maximum rotation frequency achievable in mode entrainment by applied rotating fields. It is also shown that the rotation non-uniformities caused by an EF can be used to diagnose and correct said EF in real time. Simulations using typical DIII-D data show that this can be achieved within a small number of mode rotation periods by an iterative learning control algorithm. In addition to rapidly correcting the EF, this gives access to high entrainment frequencies that would not be accessible otherwise, and paves the way to rotational stabilization and improved mode control.