Table of contents

The scaling of plasma turbulence, turbulent particle flux and cross-phase with shear in the edge of the TEXTOR-94 tokamak is obtained from measurements from fast scanning probes and compared with various existing analytical theories. It is found that the scaling can be expressed as a second order polynomial and that the cross-phase plays a key role in the suppression of particle flux. The variable rate of shear, kept below the value required to produce a low to high particle confinement transition, was obtained by changing, on a shot to shot basis, the voltage applied to an electrode introduced 4 cm into the plasma.

The tight aspect ratios (typically A≈1.4) and low magnetic field of spherical tokamak (ST) plasmas, when combined with densities approaching the Greenwald limit, provide a significant challenge for all currently available auxiliary heating and current drive schemes. NBI heating and current drive are difficult to interpret in sub-megampere machines, as in order to achieve suitable penetration into the plasma core, fast ions have to be highly suprathermal and, as a result of the low magnetic field, can be non-adiabatic (i.e. non-conserving of magnetic moment µ₀). The physics of NBI heating in START is discussed. The neutral beam injector deployed on START was clearly successful, having been instrumental in producing a world record tokamak toroidal beta of ≈40%. A fast ion Monte Carlo code (LOCUST) is described that was developed to model non-adiabatic fast ion topologies together with a high level of charge exchange loss and cross-field transport (present in START due to an envelope of high density gas surrounding the plasma). Model predictions compare well with experimental data, collected using a scanning neutral particle analyser, bolometric instruments and equilibrium reconstruction using EFIT. In particular, beta calculations based upon reconstruction of the pressure profile (by combining measurements from Thomson scattering, charge exchange recombination spectroscopy and model predictions for the fast ion distribution function) agree well with beta values calculated using EFIT alone (the routine method for calculation of START beta). These results thus provide increased confidence in the ability of STs to sustain high beta high confinement H mode plasmas and in addition indicate that the injected fast ions in collisional START plasmas evolve mainly due to collisional and charge exchange processes, without driving any significant performance degrading fast particle MHD activity.

Sawtooth inversion radii and profile peaking factors of a large variety of ohmic and ECH heated L mode plasmas, including elongations up to 2.6 and triangularities between -0.5 and 0.75, have been investigated in the TCV tokamak. In ohmic plasmas, normalized inversion radii and electron temperature profile peaking factors (corrected for sawtoothing effects) depend solely on the parameter ⟨j⟩/q₀j₀, irrespective of plasma shape. With ECH this parameter remains the main scaling parameter. Density profiles are well described as functions of poloidal flux, in agreement with turbulent equipartition theories. Parameter conversions are also provided that allow the observed scalings to be expressed using the conventional scaling variables q₉₅, δ₉₅ and κ₉₅.

The radial electric field (E_r) properties in LHD have been investigated to give guidance towards improved confinement with a possible E_r transition and bifurcation. The ambipolar E_r is calculated from the neoclassical flux on the basis of analytical formulas. This approach is appropriate to clarifying the E_r properties over a wide parameter range in a more transparent way. A comparison between calculated E_r and the experimental values has shown a qualitatively good agreement, for example, for the threshold density for the transition from the ion root to the electron root. The calculations also reproduce well the experimentally observed tendency that the electron root is possible by increasing temperatures even for higher density and the ion root is enhanced for higher density. On the basis of the usefulness of this approach, calculations over a wide range have been performed to clarify the parameter region where the E_r transition and bifurcation are possible. This gives a comprehensive understanding of the parameter region of interest, which is valuable for the promotion of experiments towards improved confinement.

During beam injection in the DIII-D tokamak, modes with lower frequencies than expected for toroidicity induced Alfvén eigenmodes (TAEs) are often observed. The experimental `TAE' frequency is often ≈0.8 of the nominal theoretical frequency of the TAE, f<sub>TAE</sub>, while the typical frequency of beta induced Alfvén eigenmodes (BAEs) is (0.2-0.4)f<sub>TAE</sub>. An analysis is presented of an unstable discharge with a high n stability code, HINST, that includes the effect of energetic ions on mode frequency. The analysis shows that the experimental `TAE' and `BAE' could be resonant branches of the TAE and the kinetic ballooning mode, respectively.

Particle diffusion in a given electrostatic turbulence with a finite correlation length along the confining magnetic field is studied in the test particle approach. An anomalous diffusion regime with amplified diffusion coefficient is found under the conditions when trajectory trapping in the structure of the stochastic potential is effective.

Recycling and wall pumping have been studied comparing low (≈10¹⁸ m^-3) and high (≈10¹⁹ m^-3) density long duration plasmas in TRIAM-1M. The recycling coefficient of each plasma increases with time. There exist two time constants in the temporal evolution of the recycling coefficient. One is a few seconds and the other is about 30 s. The wall pumping rates of low and high density plasmas are evaluated to be ≈1.5×10¹⁶ atoms m^-2 s^-1 and ≈4×10¹⁷ atoms m^-2 s^-1, respectively. The difference may be caused by the total number of diffused ions and the charge exchange neutral flux with an energy of less than 0.7 keV. In an ultra-long discharge (≈70 min), the recycling coefficient becomes 1 or more before again decreasing below 1, i.e. the wall repeats the process of being saturated and refreshed. In high power and high density experiments, wall saturation phenomena have been observed. The discharge duration limited by wall saturation decreases with an increase in density.

An independent assessment of the physics of Ignitor is presented. Ignitor is a physics demonstration experiment with the main goal of achieving thermonuclear ignition, defined as the regime where fusion alpha heating compensates for all forms of energy losses. Simulations show that a pulse of alpha particle power of up to 10-20 MW is produced for a period of more than a few seconds. Crucial issues are the production of peaked density profiles on a timescale of several energy confinement times, the control of current penetration for the optimization of ohmic heating and sawtooth avoidance. The presence of a 10-20 MW ICRF system and the operation of a high speed pellet injector are considered essential to provide added flexibility in order to counter unexpected, adverse plasma behaviour.

Experimental results towards long sustainment of JT-60U reversed shear (RS) plasmas, where high confinement is achieved owing to strong internal transport barriers, are reported. In a high current plasma with an L mode edge, a deuterium-tritium-equivalent fusion power gain Q_DT^eq = 0.5 was sustained for 0.8 s (≈the energy confinement time) by precise adjustment of the plasma beta using feedback control of stored energy. In a high triangularity plasma with an ELMy H mode edge, a high bootstrap current fraction (≈80%) was obtained in a high q regime (q₉₅≈9), which led to the full non-inductive current drive condition. A normalized beta β_N ≈2 and an H factor H₈₉≈3.5 (HH_98y2≈2.2) were sustained for 2.7 s (about 6 times the energy confinement time). The values of β_N and H₈₉ sustained in RS plasmas were improved remarkably by operating with high bootstrap current fraction and high triangularity.

The modelling studies being performed for steady state divertor operation of the ITER-FEAT design are summarized. Optimization of the divertor geometry reveals the importance of the proper target shape for a reduction of the peak power loads. A high gas conductance between the divertor legs is also essential for maintaining acceptable conditions in the outer divertor, which receives a higher power loading than the inner divertor. Impurity seeding, which would be necessary if tritium co-deposition concerns precluded the use of carbon as the plasma facing material, can ensure the required high radiation level at acceptable Z_eff, and the divertor performance is not very sensitive to the choice of radiating impurity.

In the l = 3/m = 9 Uragan-3M (U-3M) torsatron (R₀ = 1 m, ā≈0.12 m, B_ϕ = 0.72 T, ι(ā)/2π≈0.4), an open helical divertor is realized. A hydrogen plasma with _e≈2×10¹⁸ m^-3, T_e≈0.3 keV, T_i≈0.1 keV is produced and heated by RF fields (ω≈ω_ci). The flows of diverted plasma are detected by 78 plane Langmuir probes aligned poloidally in the spacings between the helical coils in two geometrically symmetric poloidal cross-sections of the torus. In measurements of the distributions of ambipolar (e.g. the ion saturation current I_s) and non-ambipolar (e.g. the current to a grounded probe I_p) plasma flows, a strong vertical asymmetry of these distributions is observed, its main characteristics being a many-fold difference in the values of I_s in the outgoing flows in the upper and lower parts of the torus and the opposite signs of I_p in these flows, with the positive current corresponding to the larger ambipolar flow of the diverted plasma. Reversal of the direction of the toroidal magnetic field results in the reversal of the asymmetry, with the larger flux (and I_p>0) always flowing in the ion B×∇B drift direction. On this basis, it is concluded that the asymmetry is related to direct (non-diffusive) losses of charged particles from the confinement volume. This conclusion is validated by numerical modelling of thermal and fast particle orbits in U-3M, where qualitative agreement has been revealed between the calculated distribution of the angular co-ordinates of lost particles and the measured poloidal distributions of the flows of diverted plasma.

The absolute sputtering yields of D⁺, He⁺ and Li⁺ on deuterium saturated solid lithium have been measured and modelled at 45° incidence in the energy range 100-1000 eV. The Ion-surface InterAction Experiment (IIAX) was used to measure the absolute sputtering yield of lithium in the solid phase from bombardment with a Colutron ion source. The lithium sample was treated with a deuterium plasma from a hollow cathode source. Measurements also include bombardment of non-deuterium-saturated lithium surfaces. The results lead to the conclusion that the chemical state of the deuterium treated lithium surface plays a major role in the decrease of the lithium sputtering yield. Specifically, preferential sputtering of implanted deuterium atoms over lithium atoms in deuterium treated samples results in a decrease of at least 60% of the lithium sputtering yield, in the case of He⁺ bombardment. These results also demonstrate that lithium self-sputtering is well below unity and that the fraction of sputtered species in an ionic state ranges from 55 to 65% for incident particle energies between 100 and 1000 eV. Furthermore, correlation of Monte Carlo VFTRIM-3D simulations and IIAX experimental data demonstrate that the surface composition has a one to one ratio between deuterium and lithium components.

Data are presented that are part of a first step in establishing the scientific basis of magnetized target fusion (MTF) as a cost effective approach to fusion energy. A radially converging flux compressor shell with characteristics suitable for MTF is demonstrated to be feasible. The key scientific and engineering question for this experiment is whether the large radial force density required to uniformly pinch this cylindrical shell would do so without buckling or kinking its shape. The time evolution of the shell has been measured with several independent diagnostic methods. The uniformity, height to diameter ratio and radial convergence are all better than required to compress a high density field reversed configuration to fusion relevant temperature and density.