The Advanced Limiter Test (ALT) project is the focus of a fruitful and intense International Energy Agreement collaboration on TEXTOR. The pump limiter is a mechanical boundary that is laid out for taking the full heat load of TEXTOR, namely 8 MW (assuming 2 MW radiated power)for 10 s, and provides a pumping efficiency of at least 5% of the working gas. This layout is adopted from the requirements of a fusion reactor: It is mandatory to remove both the full power that is convected to the limiter or divertor and the helium ash that is generated in the fusion process. In order to obtain pumping for all gases, the ALT-II is equipped with turbomolecular pumps. A short description of ALT-II is given, and the power and particle fluxes to the limiter surface and into the exhaust scoops are discussed. Requirements of the helium removal rate for a reactor and relevant measurements are discussed, and particle removal and the power distribution to the limiters are treated. Related topics of the ALT-II program were hydrogen recycling and the measurement of turbulence-induced anomalous particle transport in the plasma edge.

Particle exhaust studies have been carried out with the pump limiter ALT-II in the TEXTOR tokamak, under ohmic conditions as well as with NBI and with ICRF auxiliary heating, and the pumping effectiveness is shown to meet the requirements for a fusion reactor. Quantitative measurements of Dα emission, made with a CCD camera, have been used to determine the particle efflux from the plasma. Roughly one third of the Dα emission occurs in a diffuse `halo' that surrounds the limiter belt. The particle confinement time is less than the energy confinement time by a factor of typically 4. Modelling in 2-D of plasma and neutral flows in the TEXTOR boundary has been performed. The source of D+ ions can be related to the Dα emission by a factor that is found to depend on the location of the emission and on the discharge density. The predicted total Dα emission agrees with the measurements within a factor of about 2. Pumping of ALT-II allows for density control; with NBI, the density can be increased well beyond the ohmic limit without the discharge ending in disruption. The plasma particle efflux and the pumped flux both increase with density as well as with heating power. The exhaust efficiency is typically ∼2%, with the highest values observed in high density NBI discharges. Higher exhaust rates are observed with NBI than with ICRF. Plasma and neutral flows in the ALT-II scoops have been simulated, making use of a simple plasma model. The scoop may be viewed as a non-linear amplifier of the plasma particle flux; the amplification is found to range from about 2 to 3 for most cases. Flow reversal in the scoop is found in some of the NBI cases and particularly in the highest density case.

Electrostatic biasing experiments using the Advanced Limiter Test (ALT-II) pump limiter in the TEXTOR tokamak have been carried out with the dual goals of: (a) improving the core plasma confinement in the tokamak and (b) enhancing the performance of the pump limiter. The fully toroidal belt limiter has been biased during both ohmic and neutral beam heated discharges. Both polarities of bias have been applied up to a maximum of +or-500 V with no evidence of impurity accumulation in the central plasma, although applying either polarity of bias to the limiter increases recycling from both the limiter face and the vacuum vessel liner. This in turn results in an increase of the central density. The application of a negative bias to the limiter produces a barrier to radial particle transport in the region between the limiter and the wall. This barrier is not observed in either the no bias or the positive bias case. Neither polarity of limiter bias affects the central plasma energy confinement, apparently because the electric field structure producing the radial barrier is outside the limiter tangency radius. The enhanced recycling, coupled with high edge density, increases the radiated power from the plasma edge and may lower the power flux to the plasma facing surface of the limiter blade. In the case of positive limiter biasing, the pressure in the pumped plasma collection scoops of the limiter increases by approximately 20%, corresponding to a similar increase in the particle removal rare of the pump limiter. The increase in the particle removal rate appears to result from a lower edge electron temperature. This is consistent with the observation of an increase in edge radiated power

The compatibility of radiative edge cooling by neon injection and He exhaust with the pump limiter ALT-II is studied on TEXTOR. It is demonstrated that in plasmas with strong auxiliary heating (2 MW NBI) and with the highest average electron densities (n(e)BAR = 5.5 X 10(19) m-3) the effective confinement time tau(P)* for He has a minimum. This good pumping performance is maintained even for cases in which up to 90% of the heating power is radiated from the plasma boundary (cold radiative edge). The processes inside the scoops of the pump limiter (neutral particle transport, re-ionization) and the variation of particle confinement in the main plasma dominate this behavior.

Experiments have been carried out in the TEXTOR tokamak to investigate the performance of the pump limiter ALT-II at the highest achievable core densities. Neutral beam heating allows the density to be extended beyond the ohmic limit. The use of one beam raises the line-averaged density limit by approximately 50%. Results from discharges with co- and balanced injection are shown. Plasma fluxes to the pump limiter increase with density. Up to 4.5 X 10(21) ions/s (720 A) are collected in the eight segment belt, with pumped particle fluxes of up to 3 X 10(21) atoms/s (60 mbarl/s). Exhaust efficiency determination requires knowledge of the efflux of plasma from the core, which can be estimated from SOL profiles and from D(alpha) emission. Both exhaust efficiency estimates increase with density. Experimental results are discussed in terms of predictions from the neutral transport code EIRENE.

Thermographic measurements at the ALT-II pump limiter on TEXTOR have been performed with the following results. (1) The design of the toroidal pump limiter provides a rather uniform temperature distribution and prevents an overheating of the graphite cover tiles. (2) The local measurements of the convective energy flux to the limiter (IR scanner) and of the radiated power (bolometer) show normally a good power balance; this indicates that the power is leaving the plasma toroidally symmetrically. (3) From the analysis of the temperature distribution in poloidal and toroidal direction along the ALT-II tiles a power decay length of 5-7 mm is derived. (4) The ion temperature at the last closed flux surface is obtained from the energy flux along the magnetic field lines. At low plasma densities, the ion temperature is up to a factor of 2 higher than the electron temperature and at high densities both temperatures converge to the same value.

In the TMI tokamak current spikes into elements of the poloidal limiter were observed during disruptions. Currents flowed between limiter elements located at the outboard side and inboard side of the torus as well. Simultaneously, the variation of the plasma column vertical position is observed. The direction of this shift coincided with that of a force produced by currents between poloidal limiter elements and the toroidal magnetic field. The correlations between vertical shift, current flowing to limiters and the shift along the major radius were investigated in the small TF tokamak. During major disruptions in the TEXTOR tokamak the currents flowing between the ALT-II toroidal belt limiter and the liner were recorded.

Poloidal asymmetries in the electron density profiles and limiter diagnostics suggest that poloidal and toroidal flows exist in the SOL and edge plasma of the TEXTOR tokamak. A fast reciprocating probe located at the outer midplane provides profiles of the basic plasma parameters including radial electric field and fluctuation driven transport. Based on the probe measurements, we calculate the E(r) X B(t) drift velocities and phase velocities of the turbulence. The measurements are compared to each other. The above measurements are compared to calculations using the SOLXY code for B(t) and I(p) in normal and inverted configuration.

The detection of helium in a deuterium gas environment (both mass 4) is an important task in the field of nuclear fusion physics. To fulfill this purpose a detector has been developed which is based on spectroscopic measurement techniques. For the excitation of the helium and deuterium atoms different electron sources have been tested. Most sources had to be rejected because they either produce a disturbingly high background light level or cannot be operated reproducibly after venting or are mechanically not stable enough. The best line emitting light source has been found to be a Penning gauge. The light is collected in an optical fiber bundle, transferred away from the tokamak experiment to an area more easily accessible, split by interference filters into Dα light and the light of a strong He i line, and detected by photomultipliers. Unfortunately broad spectral lines from the deuterium molecule superimpose the strongest He i lines and have nearly the same intensity as the helium lines at a He concentration of about 5%. Therefore, some effort is necessary to deduce the partial pressures of helium and deuterium. A method is described which yields the calibration factors for the observed nonlinear pressure response of the spectral lines. The lower limit for the determination of the helium concentration presently amounts to about 1%–2%; the time resolution of the system is of the order of a few milliseconds.

Edge radial electric fields were induced in the edge of the TEXTOR tokamak by means of a polarization electrode in order to study their influence on the plasma edge profiles and its confinement. The studies include the generation of H-mode behaviour with either positive or negative polarity. Particle confinement (τp) of deuterium and of impurity ions as well as energy confinement (τE) are investigated. For positive fields which remain below the threshold for the L-H transition, an interesting regime of reduced particle confinement without noticeable energy confinement loss is found. A strong asymmetry in the edge density profiles with respect to the electric field sign is observed at these low polarization voltages. Above the threshold, H-mode behaviour with increased energy confinement and especially particle confinement can be produced with either polarity of the applied electric field. It is, however, found that, whereas the energy confinement in positive H-modes is at least as good as that in negative ones, the ratio τp/τE is about three times lower in the former case

Rapidly changing heat fluxes deposited on the limiter blades are observed during disruptions by infrared (IR) scanners. These scanners are a suitable tool for the analysis of these heat fluxes because they provide both spatial and temporal information with sufficient resolution. Several new features of the power flux to the plasma facing surfaces, during a disruption have been found. The disruptive heat flux occurs on three different time-scales. The fastest ones are for heat bursts with a duration of ⩽ 0.1 ms; several of these bursts form a thermal quench of about one millisecond duration, and some of these thermal quenches are found to occur during the current decay phase. Power flux densities of the order of 50 MW/m2 have been observed during a burst. The spatial extent of the area on which this power is deposited during a burst is larger than or equal to the size of half an ALT-II blade, i.e. about 1 m in the toroidal direction. Simultaneous measurements with two cameras show that the correlation length of a single burst is smaller than half the toroidal circumference, probably of the order of half a blade or a full blade length. This is consistent with plasma islands of low mode number. The typical heat deposition patterns at the limiter blades for normal discharges are preserved during a disruption. The magnetic structure near the plasma surface can therefore not be destroyed completely during the thermal quench. The power flux follows the field lines. However, the power e-folding length is about a factor of two to three times larger than under normal discharge conditions

Helium removal experiments were conducted in TEXTOR with a small helium self-pumping module located in a modified ALT-I limiter head. The module contained two heated nickel alloy trapping plates, a nickel deposition filament array, a Langmuir probe, flux probe, and thermocouples. The experiment examined plasma helium removal via trapping of helium ions in the deposited nickel surfaces. Such helium removal was successfully observed, with about 10% of the helium in a 10% He/D plasma being removed in a approximately 1 s period. The module was found to be compatible with overall tokamak operation with essentially no sputtered nickel entering the core plasma. The temperature rise on the ion-exposed inner trapping plate, during a plasma shot, is consistent with a local sheath potential of approximately 3kT(e). Post-tokamak test examination of the trapping plates shows helium atom concentrations in the deposited nickel consistent with the observed helium removal, and shows very small D concentrations.

The Advanced Limiter Test-II (ALT-II) is a large area toroidal pump limiter in the TEXTOR tokamak. Discrete target plates located in channels at eight toroidal positions behind the main limiter surface neutralize a portion of the plasma efflux from the core. The resulting gas is exhausted by eight external pumps. The primary experimental goals of ALT-II are aimed at power loading studies and plasma density control during long pulse (4 s), high power (6 MW) tokamak discharges. It is found that both the power and the plasma flow to the limiter are asymmetric and depend on line density. Peak neutral pressures of 0.8 m torr and removal rates of up to 0.15 torr⋅L/s per pump station are achieved in the Ohmic phase. The projected exhaust efficiency of ALT-II with full pumping is 5-10%. During ion cyclotron resonance frequency heating, the particle removal rate exceeds 0.4 torr⋅L/s per blade, and the exhaust efficiency is 4-5% for power levels up to 2.6 MW.

Many studies have shown the importance of the ratio τHe/τE in determining the level of He ash accumulation in future reactor systems. Results of the first tokamak He removal experiments have been analyzed, and a first estimate of the ratio τHe/τE to be expected for future reactor systems has been made. The experiments were carried out for neutral-beam-heated plasmas in the TEXTOR tokamak at KFA Jülich. Helium was injected both as a short puff and continuously and subsequently extracted with the Advanced Limiter Test-II (ALT-II) pump limiter. The rate at which the He density decays has been determined with absolutely calibrated charge-exchange spectroscopy and compared with theoretical models, using the Multiple Impurity Species Transport (MIST) code. An analysis of energy confinement has been made with the Princeton Plasma Physics Laboratory (PPPL) TRANSP code, to distinguish beam from thermal confinement, especially for low-density cases. The ALT-II pump limiter system is found to exhaust the He with a maximum exhaust efficiency (eight pumps) of ∼8%. We find 1< τHe/τE < 3.3 for the database of cases analyzed to date. Analysis with the International Thermonuclear Experimental Reactor (ITER) TETRA systems code shows that these values would be adequate to achieve the required He concentration with the present ITER divertor He extraction system.

Modern optical fibers, through control of the purity of the materials and the tolerances of the core and clad diameters, provide very good light transmission in the visible and near-ultraviolet regions of the spectrum. This makes it possible to use them in place of traditional optical systems without large losses in light intensity at the detectors. In addition, the same control of the quality of the fiber materials, coupled with novel jacket materials, makes it possible to use the fibers inside vacuum chambers and at elevated temperatures. A fiber-optic bundle recently installed in the TEXTOR tokamak is an example of the use of modern fiber technology. The bundle was made of 80 100-μm fibers held together with a polyimide organic material that has good outgassing specifications up to 400 °C. This fiber bundle has been used for recent measurements of the recycling in the throat region of one of the blades of the Advanced Limiter Test-II (ALT-II) belt pump limiter. Another system presently under design and testing employs individual fibers that are gold plated. These fibers are fed through holes in a vacuum blank flange and silver soldered to the flange. This system is designed to transmit the light from the strike point inside the closed divertor of the DIII-D tokamak out to a spectrometer. There, the spectral profile of the Hα line is analyzed to determine the energy distribution of the recycling particles.

The ALT-II (Advanced Limiter Test) toroidal belt pump limiter has been used for particle exhaust during discharges heated by neutral beam injection (NI) in the TEXTOR tokamak. Pumping is now available at all eight blades, whereas previously only two blades have been pumped. Results from discharges with co- and with counter-injection are presented in this paper. Large rises (~ 3 × over ohmic conditions) in collected flux are observed in the collection scoops with NI, but the electron temperature rise is modest. The plasma density near the neutralizes is substantially higher than at the scoop entrances. Total collected fluxes at the neutralizers have reached 16 × 1020 ions/s (260 A), and removal rates of nearly 7 × 1020 atoms/s (15 mbarl/s) have been measured, compared to 1.6 × 1020 atoms/s injected in a neutral beam.

A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, approximately 25 X 25 cm2 heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of approximately 50 angstrom nickel will be deposited on the two trapping plates. During a shot helium and hydrogen ions will impinge on the plates through a approximately 3 cm wide entrance slot. The helium removal capability, due to trapping in the nickel, will be assessed for a variety of plasma conditions.In support of the tokamak experiment, the trapping of helium over a range of ion fluences and surface temperatures, and detrapping during subsequent exposure to hydrogen, were measured in ion beam experiments using evaporated nickel surfaces similar to that expected in TEXTOR. Also, the retention of H and He after exposure of a nickel surface to mixed He/H plasmas has been measured. The results appear favorable, showing high helium trapping (approximately 10-50% He/Ni) and little or no detrapping by hydrogen. The TEXTOR experiment is planned to begin in 1991.

Extensive investigations of ICRF-induced effects on the edge plasma and on plasma-wall interaction were conducted on TEXTOR under different wall- and limiter as well as plasma- and heating conditions.