LAUSR

The article begins with a brief review of the achievements of Russian tokamaks in the active period of their development from 1962 to 1973, under the leadership of academician L.A. Artsimovich. During these years, the following basic issues were solved: the equilibrium problem, the MHD stability of the plasma column, and creation of the hot plasma with intense DD neutron radiation. It was shown that the ion confinement in tokamaks is close to the neoclassical model, and the electron confinement is abnormal. It improves with the increasing frequency of collisions, the opposite of the case with ions, in what is known as the alternative model of confinement along the magnetic field. Finally, the first scaling law for the energy lifetime of plasma was obtained, which accurately predicted the plasma parameters of the next generation of tokamaks (the so-called T-4 scaling). The subsequent movement in this direction (the 'Artsimovich vector') led to the creation of DT reactors with a fusion power of up to 10 MW (TFTR, JET) and to the ITER project. The main objective of the further development of tokamaks is their transition to steady-state fusion operation, which is a prerequisite for their use in industrial power generation. This makes it necessary to re-evaluate the achievements and obstacles that have to be overcome. The first limitation which thus arises is the so-called P H/S limit, which limits the value of the plasma heating power in a tokamak, as well as the discharge duration ▵t (the 'TRIAM vector') in current tokamaks (▵t ~ 1/(P H/S)1.7). Analysis of the existing experimental data shows that the most probable reason for the limitation of P H in existing tokamaks is the breakdown of the plasma sheath in the places of direct contact of the plasma with the wall. The reason for limiting the discharge duration ▵t may be the gradual accumulation of the erosion products in the contact zones of the plasma with the tokamak first wall, which can facilitate such a breakdown. Creating a closed circulating lithium flow between the first wall and plasma is proposed as the solution to the problem of accumulation of the products of the first wall erosion. Preliminary studies (appendices A and B) have shown that the undesirable accumulation of tritium in the protective lithium films can be avoided if the temperature of the wall of the tokamak discharge chamber does not exceed 400 °C.