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Controlled thermonuclear fusion. Association Euratom/KFA

Objective

In accordance with its specialized background and multi- disciplinary nature, KFA Julich as concentrated on the subject area of plasma and wall technology. The Tokamak TEXTOR was specially developed as a testing facility in this connection. The objective is the development of a wall system - and the materials suitable for it - which can withstand the intense heat and particle fluxes from the hot plasma over a long period ("Plasma Facing Components and Materials") and is also compatible with the plasma-physics requirements.
A regime of very high confinement (VH-mode) has been observed in divertor discharges in DIII-D. The VH-mode, first seen following the initial boronization of the DIII-D vessel in 1991, exhibits total energy confinement a factor of 2.5 to 3.5 greater than that predicted by the ITER89-P L-mode scaling relation. Also, confinement of thermal energy alone is greater than 1.6 times that of the JET/DIII-D H-mode scaling and in many cases has exceeded twice that amount. Some of the results of experimentation and analysis in this field, detailing some of the characteristics of VH-mode and making comparisons with the H-mode where appropriate, are presented.

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 included the generation of H-mode behaviour with either positive or negative polarity. Particle confinement of deuterium and of impurity ions as well as energy confinement were investigated. For positive fields which remained below the threshold for the L-H transition, an interesting regime of reduced particle confinement without noticeable energy confinement loss was found. A strong asymmetry in the edge density profiles with respect to the electric field sign was observed at these low polarization voltages. Above the threshold, H-mode behaviour wth increased energy confinement, and especially particle confinement, could be produced with either polarity of the applied electric field. It was found that whereas the energy confinement in positive H-modes was at least as good as that in negative ones, the ration of particle confinement to energy confinement was about 3 times lower in the former case.

Density limits and the evolution of disruptions were studied on TEXTOR for different wall materials facing the plasma surface. A systematic extension of the density limit was found over a large current range when the metallic wall was first exchanged by a carbon coated wall and later by a boron carbon coated wall. The maximum attainable density nearly doubled when metallic impurities were replaced by low Z impurities such as boron and carbon. With a neutral beam additional heating power of 3.2 MW, average densities exceeding 10E20 per cubic metre were obtained. The maximum density observed scales as the square root of the auxiliary heating power. The experimental data obtained for the different materials are presented, and ohmic and neutral beam heated plasmas are compared. The physical processes leading to the major plasma disruption and its time evolution, are also described.

In order to implement a new setup for netural beam activated impact excitation spectroscopy at the nuclear fusion experiment TEXTOR (Tokamak Experiment for Technology Oriented Research) a 2.45 GHz ECR multicusp ion source was built. Special emphasis was given to the extraction of an intense He+ ion beam. After neutralization, the slow (2.2E7 cm/s, 1 keV) helium atoms are injected into the tokamak plasma. The atom beam needs to have a small divergence.

To optimize the extraction optics, calculations concerning the multihole extraction aperture have been made, taking different ECR plasma parameters into account. The divergence shows a strong dependence on the extracted beam current density (greater than 5 mA/square centimetre) and the chosen optics. The ion source has to be operated in a pulsed mode (maximun 5 kHz, duty cycle 50%) due to experimental constraints.

Rapidly changing heat fluxes deposited on the limiter blades were observed during disruptions by infrared (IR) scanners. Several new features of the power flux to the plasma facing surfaces during a disruption were found.

The disruptive heat flux occurs on 3 different time scales. The fastest ones are for heat bursts with a duration of less than 0.1 ms; several of these bursts form a thermal quench of aobut one millisecond duration, and some of these thermal quenches may occur during the current decay phase. Power flux densities of ther order of 50 MW/square metre were 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, ie about 1 m in the toroidal direction. Simultaneous measurements with 2 cameras showed 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 therefore cannot be destroyed completely during the thernmal quench. The power flux follows thefield lines. However, the power e-folding length is about a factor of 2 to 3 times larger than under normal discharge conditions.

Measurements of the ion and electron temperature and density in an electron cyclotron discharge, by means of the rotatable double probe and a retarding field analyser, are presented. A significant resonance absorption at the second harmonic was observed. A discussion of the relevance of the results for the optimum operation of such a discharge as an ion source follows.

A thin boron film was applied to the DIII-D tokamak plasma facing surfaces to reduce impurity influx, particularly oxygen and carbon. A direct result of this surface modification was the observation of regime of very high energy confinement, VH-mode, with confinement times from 1.5 to 2 times greater than predicted by the H-mode scaling relation for the same set of parameters. VH-mode discharges are characterized by low ohmic target densities, low edge neutral pressure, and reduced recycling. These conditions have reduced the collisionality in the edge region, producing a higher edge pressure gradient and a significant bootstrap current, up to 30% of the total current.

The edge plasma properties after boronization are described, including reductions in recycling. In particular, the edge plasma conditions necessary for access to VH-mode, including the boronization process and properties of the deposited film, are discussed.

A fundamental investigation of barrier discharges has been undertaken, with a view to using this kind of gas discharge for surface treatment. As an introduction, the characteristic properties of barrier discharges are outlined, but interest centres on the microdischarges which make up the barrier discharges.

Experimental and measuring equipment used in this investigation is described. The time and space dependent behaviour of barrier discharges was studied, and the type and pressure of the gas, the geometric arrangement and the supplied voltage were varied in order to achieve a uniform distribution of the microdischarges. In the characterization of discharges, emphasis was placed on their current and voltage shapes. Discharge conditions suitable for the treatment of surfaces are discussed.

The surface reaction of thermal atomic hydrogen with polycrystalline diamond film prepared by hot filament assisted chemical vapour deposition has been investigated. Atomic beam techniques in combination with mass spectroscopic detection of the formed product molecules have been used to study hydrogen-induced erosion. When compared with amorphous carbon hydrogen films, the reaction of hydrogen with diamond films is negligible. The yield is orders of magnitude smaller and reaches maximum values below 1E-4 around 500 K. After starting the atomic hydrogen exposure, a transient behaviour is observed with 10 times higher yield, probably due to a reaction with remaining amorphous carbon.

The numerical results concerning the longitudinal ion viscosities, heat conductivities and thermal diffusion ratios in magnetised deutium-helium and hydrogen-carbon plasmas within Grad's 21 moment approximation are presented. Calculations taking full account of the finite mass ratio between hydrogen (deuterium) and impurity ion mass are compared with those for values of finite man ratio much less than 1. As expected, the largest discrepancies between exact and approximate results (with respect to the mass ratio) arise for deutium-helium plasmas. It is shown that the approximate formula for the impurity ion heat conductivity has to be extended to correct previous deficiencies.
History of Association/Laboratory

The "Institut für Plasmaphysik" was the first scientific institute of KFA Jülich, founded in the late fifties. The Contract of Association with EURATOM was signed in 1962. During the sixties the programme was focused on theta pinches. In the mid-seventies the activities were re-oriented towards plasma-wall interaction. The central facility for the experimental programme, the tokamak TEXTOR (Torus EXperiment for Technology Oriented Research), became operational in 1983 and has been recentlly upgraded by significant pulse lenght prolongation ("TEXTOR 94").

Present scientific and technical programme

The essence of the programme is plasma-wall physics and technology with emphasis on heat removal and particle exhaust under long-pulse high-power heating conditions. The limiter tokamak TEXTOR (R = 1.75 m, a = 0.5 m) is equipped with 4.0 MW ICRH and 4.0 MW NBI heating (providing a power flux density through the boundary of 25 W/cm{2}) and with a toroidal pump limiter (ALT-II). With TEXTOR 94 fusion relevant plasmas of up to about 10 seconds duration are being produced, maintained and studied. Related issues of plasma confinement, transportand modelling are also addressed, as well as the development of plasma facing materials. Further teststands for applying extremely high heat loads on first-wall components and materials, both by ions and electrons, are available.
Highlights of the TEXTOR programme : wall conditioning by boronisation and siliconisation, efficient helium removal, I-mode (improved confinement), edge radiation cooling.

Staff

Professionals : about 60
Support staff : about 80

Yearly budget (expenditure 1994): about 23 Mio ECU

Management structure

Head of Research Unit : G.H. WOLF

Collaboration with other institutions

The Association EURATOM/Belgian State (ERM/KMS Brussels) has taken full rsponsability of the heating programme in TEXTOR. The Dutch FOM Association is engaged in TEXTOR with several diagnostic equipments. These two partners and the KFA Association are ready to combine and focus their efforts in fusion oriented plasma physics by establishing a new transnational organisational structure under the name Trilateral Euregio Cluster (TEC). The joint programme is oriented towards the needs of ITER and the W7-X stellator project. Collaboration with the Universities in Belgium, The Netherlands and Nordrhein-Westfalen is promoted by the Euregional Club of High-Temperature Plasma Physics (ECPP). Strong intercontinental links are established through the IEA Implementing Agreement on TEXTOR (partners: Canada, EURATOM, Japan, Switzerland, USA).

Topic(s)

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Call for proposal

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Funding Scheme

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Coordinator

Forschungszentrum Jülich (KFA)
EU contribution
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Address

52425 Jülich
Germany

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