Techniques for cleaning and conditioning the vacuum vessel of the Advanced Toroidal Facility (ATF) and its internal components are described. The vacuum vessel cleaning technique combines baking to 150"C and glow discharges with hydrogen gas. Chromium gettering is used to further condition the system. The major internal components are the anodized aluminum baffles in the Thomson scattering system, a graphite-shielded ICRF antenna, two graphite limiters, and a diagnostic graphite plate. Three independent heating systems are used to bake some of the major components of the system. The major characteristics used for assessing cleanliness and conditioning progress are the maximum pressure attained during bakeout, the results of gas analysis, and relevant plasma parameters (e.g., time to radiative decay). Details of the various cleaning and conditioning procedures and results are presented.

The Quasi-Poloidal Stellarator (QPS) [J.F. Lyon, S.P. Hirshman, D.A. Spong, et al., 30th EPS Conference on Plasma Phys. Control. Fusion, St. Petersburg, 2003] has a very low plasma aspect ratio (R/a∼2.7) and is a new confinement approach that could ultimately lead to a high-beta compact stellarator reactor. QPS is being developed to test key features of this approach. In QPS, the modular stellarator coils and other components are inside the vacuum vessel, requiring special attention with respect to particle control. We have analyzed two specific aspects: (1) The casings of the internal coils will assume elevated temperatures during operation whereas the vacuum vessel will be at room temperature; we have analyzed the consequences for adsorbed gas layers. (2) The plasma is surrounded by a large gas reservoir that might cause problems for density control. We have constructed a model to predict plasma and neutral densities as functions of fueling, divertor baffling, wall adsorption, and particle exhaust. This model predicts that plasma and neutral densities can be controlled.

A candidate magnetic topology of the plasma boundary of the proposed compact stellarator national compact stellarator experiment (NCSX) is investigated using field-line tracing with diffusion. The required magnetic fields are obtained from a free-boundary equilibrium using the magnetic fields from external coils and bootstrap plasma currents inside the last closed magnetic surface (LCMS). These results are used to calculate the magnetic fields of the finite beta equilibria inside and outside the LCMS in a form suitable for field-line tracing. Poincare plots of field lines that diffuse outwards from starting points just inside the LCMS indicate an ergodic divertor region. Intersections of field lines with a simple limiting surface show contained patches suitable for divertor control. Undesirable regions of sharply inclined angle of intersection with the limiting surface are localized, indicating the suitability of the configuration for optimized divertor design techniques. We also discuss physics implications of field-line lengths in the divertor region.

The national compact stellarator experiment (NCSX) [EPS 2001, Madeira, Portugal, 18–22 June 2001] is a new fusion project located at Princeton Plasma Physics Laboratory, Princeton, NJ. Plasma boundary control in stellarators has been shown to be very effective in improving plasma performance [EPS 2001, Madeira, Portugal, 18–22 June 2001] and, accordingly, will be an important element from the very beginning of the NCSX design. Plasma-facing components will be developed systematically according to our understanding of the NCSX boundary, with the eventual goal to develop a divertor with all the benefits for impurity and neutrals control. Neutrals calculations have been started to investigate the effect of neutrals penetration at various cross-sections.

The purpose of the ALPS program is to identify and evaluate advanced limiter/divertor systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achieving high power density, up to 50 MW/m2, and showing compatibility of plasma-facing surfaces with plasma operation. Personnel representing a wide range of disciplines from a number of institutions are engaged in the program, where an evaluation phase of the program is planned for three years. Successful identification of promising concepts in the evaluation phase should lead to an R&D phase that includes proof-of-principle experiments.

High performance DT plasmas have been obtained using neutral beam heating with lithium (Li) conditioned graphite walls in TFTR. Values of τE > 300 ms have been obtained with neutron source rates of > 1018 n/s and nτT ≈ 1021. Also, ion temperature (Ti) > 40 keV and toroidal velocity (Vφ) > 800 km/s have been obtained. The Ti(R, t) and Vφ(R, t) profiles show strong gradients near the plasma core with ∇Vφ > 3.5 × 106/s and E × B shearing rate > 2 × 105/s realized. This strong E × B flow shear is consistent with formation of a ‘transport barrier’ in the plasma core. Measured Vφ, Ti, and carbon density, nc, profiles from charge-exchange recombination spectroscopy (CHERS) and neoclassical calculations of poloidal velocity, Vθ, are used to assess the roles of the pressure and velocity contributions to Er (or E × B) with varying Li conditioning. The profiles and gradients and resulting confinement and transport are found to vary with the amount of Li applied and the Li deposition technique. Correlations between the Vθ and Ti profiles and recycling and impurity behavior as implied from edge carbon and Dα light and Li deposition are also observed.

Significant results on steady state control of the plasma parameters have been obtained with RF systems on Tore Supra. A sustained effort is under way to improve power transmission reliability. Developments of sources and antennas are in progress for future operation on Tore Supra with purse length in tile range of several hundreds of seconds.

Recent Tore Supra results supporting the feasibility of high-confinement. long duration discharges are reported. Two-minute discharges with improved confinement have been obtained. This progress is largely due to improvements in the operating control system of Tore Supra. which now allows real-rime feedback control of global plasma parameters. A clear correlation between improved confinement and current profile shape has been established. Transport barriers for electron heat diffusion are observed in experiments where the magnetic shear is weak or negative in the central part of the plasma. These observations strongly support development of new current drive schemes for current profile control. Considerable progress concerning particle and heat exhaust has also been achieved. A new technique allowing conditioning in the presence of a toroidal magnetic field has been implemented. A vented limiter has been tested as a means of particle exhaust through collection of neutrals, and its performance is compared with that of a "classical", ion collecting throat limiter. The knowledge gained from extensive long pulse experimentation has been used to develop a new generation of plasma facing components. which will permit further development of the long discharge capability of Tore Supra.

Recent stationary improved confinement experiments with current density profile modifications in non-inductive Tore Supra operation are reported. Significant progress has been obtained by achieving long duration discharges using lower hybrid (LH) waves: (i) 2 min in the 4 T improved confinement LHEP (LH enhanced performance) regime, H-RLW = W-e/W-e-RLW = 1.6 at a current I-p = 0.8 MA, loop voltage V-1 approximate to 0.1 V, LH power P-lh = 2.5 MW (W-e, W-e-RLW are the electron energy content and Rebut-Lallia-Watkins L-mode prediction), (ii) 75 s long in a fully non-inductive LHEP regime (H-RLW = 1.4, V-1 = 0 V, I-p approximate to 0.62 MA, P-lh approximate to 3.0 MW) using a new plasma control scheme, (iii) 30 s at 1.7 MA (V-1 approximate to 0.25 V, P-lh = 2.5 MW) in the L-mode regime. MHD stability in full current drive operation and the role of weak-or reversed-central magnetic shear in the central LHEP electron temperature transition (reproduced in full current drive in a systematic manner and measured by independent electron temperature diagnostics) are discussed.Stationary high-beta(p) experiments were performed with fast-wave direct electron heating (FWEH) in a large range of operation: 42-76 MHz and B-t = 2-3.9 T. Maximum coupled power of 9.5 MW is obtained in the 48 MHz/2T configuration with good electron heating performance and improved confinement at high density (central density of approximate to 6 x 10(19) m(-3)). By increasing the magnetic shear in the gradient region, stationary improved confinement states (6 MW during 5 s, H-RLW approximate to 2.2) are reached in a reproducible manner with poloidal beta, beta(p), approaching 1 and 40% of bootstrap current.The high-bootstrap FWEH experiments have been recently combined with LHCD at reduced loop-voltage (less than or equal to 0.1 V and up to 70% of non-inductive current). A total injected power up to 7.2 MW (LH: 3.6 MW, FW: 3.6 MW) has produced stationary improved confinement with a peaked electron temperature profile (central value up to 8.0 keV) at a central density of 3.9 x 10(19) m(-3).

In view of high-power, long-pulse steady-state operation, Tore Supra has incorporated in its design the active control of heat and particles in a realistic environment. In the early experimental phase of Tore Supra, the first generation of plasma-facing components was tested, and these tests provided much physics and technological information and illuminated various operational difficulties. In particular, these experiments revealed the weakness of the graphite-to-metal brazing process originally adopted for actively cooled high-heat-flux components. Consequently, a new inner-wall technology was developed in 1994 and is to be tested in 1995-1996 with a totally rebuilt 40-deg toroidal sector. A carbon-fiber-reinforced carbon-metal compound is based on the newest brazing technology and rigorous quality control. Components such as the toroidal pump limiter and the guard limiters of plasma-heating antennas are being developed in the same way. For structures where brazing is difficult, boron carbide-coated components have been developed and installed in Tore Supra. For lower heat fluxes, a bolted concept has been designed and tested. The influence of inner-first-wall misalignment in Tore Supra on the power exhaust limitation of brazed components has been studied.Results from the technological development for the different power exhaust systems and the associated experimental knowledge obtained during plasma operation in Tore Supra are presented.

The combination of r.f. waves in the lower hybrid (LH) and ion cyclotron frequency ranges offers a versatile and efficient way of heating tokamak plasmas while controlling their transport properties and magnetohydrodynamic stability through the control of the current density profile. Experimental and theoretical studies on the applications of such plasma waves have been carried out on Tore Supra during recent years and are reported here.The LH system coupled up to 6.5 MW during 2 s at 3.7 GHz through two multijunction launchers. In the longest plasma shot, the total injected LH energy reaches a record value of 170 MJ during a 62 s LH pulse, at a power level of 2.8 MW, corresponding to an average power density of 17 MW m(-2). The ion cyclotron resonant frequency (ICRF) system (35-80 MHz) is composed of three resonant double-loop antennae. Up to 4 MW have been coupled with a single antenna, allowing a record power density through the Faraday screen of 16 MW m(-2) to be reached. 30 s steady-state r.f. pulses have been obtained with up to 54 MJ delivered to the plasma. One of the major observations has been the transition to the so-called ''stationary lower hybrid enhanced performance (LHEP) regime'' (I-p = 0.8 MA; n(e0) = 2.8 x 10(19) m(-3); P-LH = 3.2 MW) in which the (flat) central current density (q(0), approximate to 2) and (peaked) electron temperature profiles (T-e0 approximate to 6-8 keV) are fully decoupled. This regime exhibits a significant improvement of the global confinement (40%) owing to the increase in l(i), i.e. in the magnetic shear in the outer half of the discharge, supplemented by a large reduction in the electron thermal diffusivity in the central zone where the magnetic shear vanishes because of the slight off-axis character of the LH power and current deposition. TRANSP analyses show that LHEP plasmas provide access to the second ballooning stability regime. At higher current and density (I-p = 1.5 MA; n(e0) = 6 x 10(19) m(-3)), ICRH stabilization of sawteeth (4 MW) combined with lower hybrid current drive (LHCD) current profile modifications has allowed to extend the stabilized phase for up to 1 a with 3.4 MW of LH power, the duration of sawtooth-free periods increasing with increasing LH power. The dynamical properties of fast electrons during LHCD have been investigated recently on Tore Supra through power modulation experiments. It is shown that slowing down always predominates and, from the long-time evolution of the hard X-ray emission, the radial diffusion rate of the fast electrons is estimated to be 0.1-0.3 m(2) s(-1).Theoretical developments have focused on the modelling of LHCD and also on fast wave current drive (FWCD) and fast wave heating. The effect of intrinsic stochasticity on the propagation of LH waves is discussed and a fully developed statistical theory of stochastic wave diffusion and multipass absorption, with applications to Tore Supra through a wave diffusion-Fokker-Planck (WDFP) numerical code, is briefly presented. This model provides a simple explanation for the temperature dependence of the LHCD efficiency in small tokamaks. The ion cyclotron resonant heating (ICRH) full-wave code ALCYON has been upgraded to compute the power and current deposition profiles from direct electron absorption of the fast wave (electron Landau damping-transit time magnetic pumping). The code has been used to study FWCD in Tore Supra, the Joint European Torus and the International Thermonuclear Experimental Reactor.Finally a new concept of an efficiently cooled reflector antenna for LHCD applications in a steady-state reactor is briefly described.

The link between the current profile and the confinement is studied, involving various regimes: high power minority ion cyclotron resonant heating, high power lower hybrid current drive, fast wave direct electron heating and current drive and pellet enhanced performance. It is shown that the electron heat diffusivity decreases when the magnetic shear increases in the confinement zone and/or when it decreases in the plasma center.

With Lower Hybrid Current Drive (LHCD), very long pulses can be obtained in TORE SUPRA. Volt-second saving studies by LHW assisted current ramp-up, current drive and electron heating are first discussed. A 12 seconds stationary discharge with improved energy confinement time (40 %) by peaking the current profile is presented with a central electron temperature up to Te(o) approximately 10 keV . A 1 minute flat top time duration discharge has been achieved with a plasma current I(p) = 1 MA : 80 % of this current is driven by LHW. During this discharge the core density is fed continuously by deuterium gas puff indicating that the inner first wall is still pumping all along the shot. Finally results concerning the interaction of LHW with high density plasmas (peak density approximately 10(20) m-3) and 3 MeV fusion protons are given.

In order to do consistent scrape-off-layer plasma and neutral transport calculations, the 2D fluid code B2 has been externally coupled to the neutral transport code DEGAS for DIII-D. The coupling procedure is similar to simulations done for TFTR (Tokamak Fusion Test Reactor), Tore Supra, and CIT (Compact Ignition Tokamak). An `average' source approach is utilized to allow convergence between the two codes. Initial comparison of plasma quantities between the coupled code set and the B2 code alone shows that a colder, denser plasma may exist at the divertor targets than predicted by the B2 code with its internal recycling model

Particle control experiments in Tore Supra have shown that the change in core plasma density with pumping at the outboard pump limiter amounts only to a small fraction of the total number of particles extracted by the outboard pump limiter system. To develop a model for the core-wall particle exchange at the level of individual basic physical processes, the particle exchange between the core plasma and the wall has been modeled in detail for a series of Tore Supra discharges. Core and scrape-off layer particle balance and wall diffusion calculations were performed using a radial transport code and a 1D wall diffusion code. Core-wall particle exchange for the case of the outboard pump limiter is found to be mediated by the localized charge exchange flux in the near limiter region. The wall particle efflux required for detailed balance does not match that calculated to be available from the near limiter wall region unless some local heating of the wall is assumed.

A novel, electrically biasable, semiclosed divertor was installed and operated in the DIII-D lower outside divertor location. The semiclosed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. Electrical bias controls the distribution of particle recycle between the inner and outer divertors by E&oarr;×B&oarr; drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduces the sensitivity of plenum pressure to separatrix position. In particular, E&oarr;×B&oarr; drifts in the DIII-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density