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The main aspects of radiative plasma experiments carried out in the LH-heated TdeV tokamak are summarized. D 2 , N 2 and Ne puffing, and divertor biasing, are assessed in terms of their ability to increase edge and divertor radiative losses and reduce neutralizer plate heat loads. Detailed patterns of radiation and plate power deposition are compared. Combined N 2 and D 2 puffing gives the best results. Biasing enhances divertor radiation but also causes an increase in power conducted and convected to one of the divertor legs. The improvements obtained are still insufficient due to disruptive instabilities that appear at high impurity puffing rates. The instability is triggered at the edge with N 2 and in the core with Ne.

We have studied the retention of a recycling impurity, neon, and a wall-pumped impurity, nitrogen, in the closed and cryopumped divertor of the TdeV tokamak, under a variety of heating, puffing, pumping and biasing conditions. Retention times were deduced from either the decay time or the rise time of the plasma impurity content, measured spectroscopically, following either a short puff or a steady injection. For both neon and nitrogen, the compression ratio increases rapidly with the main plasma density. Retention is not degraded by plasma detachment. In density scans, other conditions being kept constant, the compression ratio for neon is found to grow with the divertor neutral pressure as ∼ P D 1.5 . However, by puffing, pumping or biasing, retention and P D can be varied quite independently of each other.

The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.

Experimental data, analysis and simulations are used to describe the physics of divertor detachment and He exhaust under detached conditions in TdeV. With increasing density, the plasma is found to detach progressively from the outboard divertor plates with a marked reduction of the ion flux to the plates, the generation of a pressure gradient between an ionization front and the target plate, and strong cross-field transport in the divertor. Local interactions between the divertor plasma and the plates are described, with evidence for carbon sputtering and molecular processes near the divertor plates. Divertor exhaust and retention continue to increase through detachment and He exhaust is not affected although the divertor He enrichment remains low but constant at about 0.2. A moderate density of seems to be sufficient both for efficient peak power load reduction at the divertor plate and good He exhaust through the divertor. Simulations of the edge and divertor plasmas using the B2/EIRENE and DIVIMP codes give reasonable agreement with the measurements and indicate possible divertor geometry improvements

The particle confinement time of ohmic double-null discharges in Tokamak de Varennes (TdeV) is determined by two different techniques, the conventional method and a new technique based on the temporal decay of the total core population following the injection of a gas puff. Both methods show a confinement time increasing with density up to a maximum of 13 ms at and decreasing at larger densities, with very little dependence on plasma current. Particle transport is analysed using fast gas puffing and Abel inversion of the seven-chord submillimetre (SMM) interferometer together with the source profiles determined by measurements. The incremental transport coefficients are obtained by testing the standard form of the particle flux function against the data during the transitory period towards equilibrium. Both perturbed diffusion and convection coefficients are found to vary approximately as the inverse of the density and almost proportionally to the plasma current. The equilibrium transport coefficients are then deduced from the experimental equilibrium density profiles and the measured incremental coefficients using a transport model developed from the data. The model is finally used to predict confinement times to be compared with experiment. The effect of divertor plate biasing on transport is also discussed.

The isotope ratio RI=D/(H+D) was measured during ohmic discharges in the magnetically diverted tokamak TdeV, using spectroscopy (Halpha and Dalpha ), gas analysis (H2, HD and D2), mass analysis of charge exchange neutrals, and detection of H and D in a collector probe, in the course of four different experimental phases: (I) 'dynamical' isotope switchover of the fuelling gas during a discharge, (II) shot-to-shot evolution with pure D2 fuelling, (III) same with walls conditioned by D2 glow discharge, and (IV) 'dynamical' switchover combined with divertor biasing and pumping. The different values obtained for RI are compared and analysed in terms of the particle recycling properties of the device. The dynamical scenario (phase I) revealed a lack of equilibrium between the edge and core isotopic composition that is an indication of the degree to which the particle flux is amplified by recycling in the vicinity of surfaces. The phase II results were fitted to a model which yielded an estimate of the magnitude of the wall inventory available for recycling. In phase III, the effectiveness of conditioning was assessed. Phase IV revealed an even more pronounced disequilibrium than phase I, and demonstrated the power of divertor biasing and pumping in exhausting the wall inventory and controlling the isotope ratio.

Experiments in which the divertor plates are biased with respect to the grounded vacuum chamber were conducted in the Tokamak de Varennes, with ohmic discharges and boronized walls. A comprehensive set of diagnostics was used and those effects relating to plasma-surface interactions and their consequences, namely recycling and impurity production, are reported here. The main conclusion is that the bias voltage actively controls the balance between the particle fluxes to the walls and the diverters, with a weak influence on the confinement. The quantities associated with plasma surface interactions, i.e. the impurity influx, the plasma contamination and the global recycling coefficient, increase as a function of the voltage applied between the plates and the wall. Typically, the quantities of interest vary by a factor of two or more in the range from -240 to 290 V. The poloidal fluxes to the upper or lower divertor are enhanced by negative or positive biasing, respectively, in agreement with the sign of the E*B drift. For negative biasing the reduction in the wall flux and increase in the poloidal flux combine to improve greatly the divertor efficiency: the divertor pressure and the impurity retention time are increased by factors of up to 5 and 8, respectively. The above effects, together with the SOL density profiles and the biasing current-voltage characteristic, are consistent with a model in which the radial transport in the SOL is limited by the ion mobility; a mobility of 0.04 m2/V.s is found. Besides allowing active density and impurity control, it is suggested that biasing could facilitate the attainment of a radiative divertor by injection of well retained impurities in the divertor

Electrically insulated divertor plates are used on TdeV (Tokamak de Varennes) [18th EPS Conference on Controlled Fusion and Plasma Physics Berlin (European Physical Society, Petit‐Lancy, 1991), Vol. 15C, Part I, pp. 1–141] to produce various biasing configurations, which can be decomposed into two basic modes. Plasma biasing, with a radial electric field Er in the scrape‐off layer (SOL), is most promising for divertor applications. The Er field is produced with a particular divertor plate geometry, causing a nonambipolar radial current and a particle flow in the Er×BT direction, toward one of the divertors (the active divertor). The pressure and impurity retention in the active divertor are shown, in the Ohmic regime, to be strongly increased by biasing. He exhaust through this divertor is increased by a factor of almost 3 with modest biasing voltages and currents scalable to larger devices. Biasing also modifies the power repartition between the divertors, with the active divertor also receiving a larger...

Preparation of the internal walls of tokamaks by plasma enhanced chemical vapour deposition (PECVD) of boron containing films has now been implemented on several machines since its development on TEXTOR. More recently, such films were deposited on the internal walls of TdeV using not only this procedure but also two new approaches: solid target boronization (STB) which consisted in inserting a low-density boronized carbon-carbon (C-C) composite into the tokamak plasma and TMB fuelling where trimethylboron was used as fuelling gas during the plasma discharge. These approaches resulted in a rapid shot to shot improvement of important parameters such as the volume averaged resistivity and radiated power over the first dozen shots when the boron source is present. Typically, the resistivity is reduced from ∼4.0×10 −7 to ∼2.5×10 −7 Ωm, comparable to the resistivity obtained with PECVD. The radiated power relative to the ohmic power is reduced by a factor of 2 from 20 to 10%. When the boron source, present during STB or TMB fuelling, is removed however, these plasma parameters start increasing. Within a few tens of shots, they have reverted to their preconditioning values, a situation which requires hundreds of shots after PECVD.

Two techniques are used to monitor the flow in the scrape-off layer of Tokamak de Varennes (TdeV); one is based on a new multipin Langmuir/Mach probe called Gundestrup while the other depends on the measurement of the upstream/downstream asymmetry of the power absorbed by a test limiter inserted into the plasma edge. Gundestrup has been used to measure the components of velocity parallel and perpendicular to the magnetic field as a function of the radial electric field. Both components vary linearly with the radial electric field and inversely as the magnetic field ( U ‖ ∝ E r / B θ and U ⊥ ∝ E r /B ). The pattern of power deposition on the test limiter implies that the flow is in the same direction as that measured by Gundestrup and the e-folding length for the power deposition is in agreement with Gundestrup measurements of temperature and density. The test limiter observations indicate that the flow reverses just inside the separatrix.

A new boronization technique, called solid target boronization (STB), has been developed and successfully applied in the Tokamak de Varennes. For this technique, a boronized carbon-carbon composite material is used as a source of boron and is exposed directly to confinement plasmas in order to induce boron emission due to thermal evaporation and sputtering. The STB technique can provide real-time boronization effects by implementing self-reconstructing boron-carbon coatings. A significant decrease in impurity radiation due to oxygen as well as several metals has been observed. Also, the fuelling rate increases considerably during STB, which is attributed to codeposition of carbon, boron and hydrogen

In previous tests the maximum observed TiC coating thickness which showed good resistance to 0.7 s thermal shocks was 400 μm. For this thickness, only microcracks perpendicular to the surface (segmentation) were observed, whereas for thicker ones, spalling inside the coatings occurred. With a combination of substrate surface modification (macroroughening or spraying a bondcoat) and preheating to 375°C before the thermal shocks, it has been possible to completely avoid the delamination during heating and to promote the resistance to delamination on cooling. This allows a doubling of the thickness of thermal shock resistant coatings up to 1mm. Hemispherical coated limiters were tested in TdeV (Tokamak de Varennes) with plasma currents of 210 kA and have absorbed 20 kW during 1.2 s associated with a power deposition factor of about 10 MW m −2 s 0.5 .

Bias voltages applied to the divertor plates in TdeV (Tokamak de Varennes) permit fine control of several main plasma parameters, including the poloidal rotation velocity near the separatrix, microturbulence, the sawtooth period and the heat pulse propagation speed. Biasing also improves the divertor efficiency in either the top or bottom closed divertor chamber depending on polarity, in agreement with E × B drifts. Negative biasing reduces carbon and CO fluxes from the wall, the loop voltage and the X-ray emissivity, all indicative of decreasing impurity contamination.

Experimental results obtained on the Tokamak de Varennes during a series of reproducible ohmic discharges are presented. Profiles of basic plasma parameters are constructed and compared with theoretical predictions. In particular, the measured plasma resistivity agrees with the neoclassical scaling rather than with Spitzer resistivity. The study of electron-density fluctuations indicates a linear dispersion relation with a propagation velocity of 3.0 × 104 cm s−1. Particle transport investigations are initiated, giving experimental diffusion and convection coefficients across the plasma radius for electrons and impurity ions.

A vertical slice of a tokamak plasma is imaged onto the entrance slit of a spectrometer, giving at the exit plane both the spectral (horizontal axis) and the spatial (vertical axis) distributions of the collected light. An intensified CID (charge injection device) camera is mounted at the exit plane to record the spatially and spectrally resolved emission profiles of low‐Z impurities, with a temporal resolution of 1/60 s. The large number of lines of sight allows for inversion of chord‐integrated signals to obtain local emissivities. Profiles from several ionization stages of carbon and oxygen are compared to the predictions of the impurity transport code MIST (multi‐ionized species transport) to deduce impurity diffusion coefficients.

Plasma contamination due to the generation of impurity molecules has been studied by mass spectrometry and by visible emission spectroscopy in the Tokamak de Varennes. The dominant effects are carbon monoxide formation, which is correlated with the residual water vapor pressure in the vacuum chamber, and the formation of C{sub 1}, C{sub 2}, and C{sub 3} hydrocarbons. The measured molecular fluxes are sufficient to account for a large part of the plasma impurity content. Visible spectroscopy indicates that the plasma is significantly affected by these chemical impurity sources. The molecules appear to originate mainly from the stainless steel walls rather than from the graphite limiters.

A system consisting of magnetically shielded ion gauges and a differentially pumped quadrupole mass spectrometer is described. It provides time‐resolved measurements of the neutral gas fluxes during discharges in the Tokamak de Varennes. A response time of 15 ms for hydrogen and ≊50 ms for impurity gases was obtained by a judicious choice of distance (to avoid magnetic disturbances), conductance, and pumping speed. An interface between the spectrometer and the data acquisition system of the tokamak allows fast (ms) selection of the mass and real‐time (≊10 ms) adjustment of the gain to optimize the signal‐to‐background ratio. The usefulness of this diagnostic is illustrated by its use in three experiments: propagation of a plasma density perturbation following pulsed (5 ms) gas puffing, exchange of the discharge gas from H2 to He, and chemical impurity (CO, CO2 ) production during discharges at the density limit and qa≊2. The correlation of low impurity fluxes with absence of disruptions in these last shot...

We report the first results of nondisruptive, central fueling of a tokamak by the injection of an accelerated spheromak compact toroid (CT). Interferometry measurements indicate central fueling of the tokamak plasma on a fast time scale (0.5 ms), with more than 50% of the CT mass used for plasma fueling. The tokamak particle inventory increased by more than 30% without disruption.

The Tokamak de Varennes has an auxiliary limiter facility which allows various materials to be exposed to plasma discharges. Thick, plasma-sprayed TiC coatings (200–500 μm) have been evaluated The factor which limits the thickness is presently the thermal shock resistance. Samples whose stoichiometry and micro-structure were analysed have been pre-tested using a 15 kW electron beam system. The results were used to optimize the deposition of coatings on hemispherical heads which were tested in the tokamak. With plasma currents of 200 kA, the heads have absorbed 30 kW during 0.7 sec leading to a power deposition of about 15 MW/m2. The coatings tested so far indicated an optimum thickness of about 400 μm which is still too thin for long-term operation. For this thickness, only microcracks (segmentation of the surface) are observed. For thinner coatings, the Inconel substrate was melted, whereas for thicker ones spalling inside the coatings was observed. No contamination of the core plasma by Ti has been observed, although the production of Ti from the surface has been observed, and Ti has been detected on collector probes placed on the vessel wall. AES evaluation of the exposed surface of a TiC flake has shown no significant difference in Ti and C content when compared to the reverse (unexposed) side.