The SMART project
Low aspect ratio tokamaks, commonly called Spherical Tokamaks (ST), constitute an attractive path to a fusion reactor or volumetric neutron source due to their high power densities at relatively modest costs and dimensions. Experimental observations as well as dedicated stability calculations indicate that ST can achieve plasma pressures (βT) and bootstrap currents well above the values obtained in conventional tokamaks overcoming some of the main challenges of the tokamak route towards a commercial nuclear fusion energy source.
The PSFT team is currently building a SMall Aspect Ratio Tokamak (SMART) at the Universiy of Seville. The main goal of the SMART tokamak is to train the next generation of fusion engineers and physicists as well as to contribute to key issues towards a fusion power plant with novel, and revolutionary approaches, that can sometimes, not be implemented in other large devices. The SMART tokamak will have a major, and minor, plasma radius of R=0.4m and a=0.25m respectively with a toroidal magnetic field at the magnetic axis Bt<1 T and a plasma current Ip<300 kA.
Vacuum Vessel Design
SMART is a compact device with a plasma major radius (R) of 0.4 m, plasma minor radius (a) of 0.25 m, an aspect ratio (A) less than 2 and an elongation (k) of more than 2.
SMART, for its first operation phases, will be equipped with 4 poloidal field coils, 4 divertor field coils, 12 toroidal field coils and a central solenoid. The heating system comprises of a Neutral Beam Injector (NBI) of 600 kW and an Electron Cyclotron Radiofrequency Heating (ECRH) of 6 kW for pre-ionization.
The vacuum vessel is one of the main components of the tokamak, and it has the purpose of giving a physical confinement to the plasma and keeping the required level of vacuum needed for the operation of the machine. The overall height of the vessel is 1730 mm and an outer diameter of 815 mm with a mass of almost 3 tons of AISI 316 L stainless steel. It will be designed with customized rectangular ports and commercially available circular ports mainly used for diagnostics, maintenance purposes and connection to the Neutral Beam Injection system and to the Vacuum Pumps. Ports are located on the upper/lower lids of the machine and on the body of the vessel. In order to reinforce the structure, ribs are needed on the body and on the top and lower lid. They do help the vessel maintaining the its physical structure but they do also have the function of housing the supports for the coils.
The design of the vacuum vessel has been done using a combination of physics based codes and finite elements analysis taking into account the combination of the pressure differences and the electromagnetic forces that arises between the interaction of the eddy currents and the magnetic fields. The design has been developed in compliance with the standards imposed by the international regulations for mechanical components.
The study of plasma equilibria is a fundamental aspect of tokamak physics, enabling not only a prediction of the geometric properties of the confined plasma but also a consideration of potential magnetohydrodynamic (MHD) instabilities and associated changes in particle transport. Highly-shaped equilibria, those exhibiting high elongation or triangularity, enforce high shear flows and altered edge gradients, influencing the behaviour and growth of instabilities. Studying how these plasma instabilities respond to changes in the plasma equilibrium shape represents an ongoing area of research.
A plasma in stable equilibrium represents a combined plasma current density distribution and magnetic topology such that the outward plasma pressure is in balance with the magnetic pressure confining it. For a 2D axisymmetric case, this force balance can be computed by the Grad-Shafranov equation, employing appropriate plasma boundary conditions set by the vessel and coil geometry. The final equilibrium therefore exhibits a magnetic topology that is a combination of vacuum magnetic field obtained from the coils and the magnetic fields self-generated by plasma.
Employing state-of-the-art numeral methods[1-3] it is possible to design and optimize plasma forming (PF) coil current configurations that produce a wide range of plasma equilibrium shapes within a given set of vessel wall and coil current boundary conditions. Further, the stability of each equilibrium state can be assessed through examining its response to applied translations or external perturbations.
 P. F. Buxton et. al., On the design and role of passive stabilisation within the ST40 spherical tokamak, Plasma Phys. Control. Fusion 60, 064008 (2018).
 P. F. Buxton et. al., Merging compression start-up predictions for ST40, Fusion Eng. Des. 123, 551-554 (2017).
 J. B. Lister, et. al., Plasma equilibrium response modelling and validation on JT-60U, Nucl. Fusion 42, 708–724 (2002).