[en] The surrounding material walls in fusion devices must fulfil three important tasks:provide high vacuum conditions necessary to provide clean fusion plasmas absorb the power produced by the α-particles in the fusion processes and injected by auxiliary heating enable the exhaust of the helium ash by thermalisation of the helium plasma ions on material surfaces in the vicinity of helium pumps.The interaction of the plasma with the surrounding wall surfaces (PSI: plasma surface interaction) is therefore a necessary condition for fusion devices and not to avoid. In the plasma wall interaction a variety of bulk material and surface processes are involved on one side together with various special processes in the near surface plasma region on the other side. They can modify the properties of the boundary and main plasma in a feed back like behaviour. A prominent example is the release of impurities from the walls by plasma particle impact which increases the energy loss of the plasma by radiation and reduces thereby the particle fluxes to and impurity release from the walls

[en] Helium has been implanted at certain temperatures between 800 and 1250⁰C into single and polycrystalline Ni-samples with implantation depths between 15 and 90 μm. Simultaneously the helium reemission from the sample is measured by a mass-spectrometer. It has been shown that the time dependence of the observed reemission rate is governed by volume diffusion of the helium. Measuring this time dependence as a function of temperature the helium diffusion constant has been determined. The He-diffusion is interpreted as a interstitial diffusion hindered by thermal vacancies. Depending on the implantation depth more or less of the implanted helium remains in the sample and forms large helium bubbles. (orig./GSCH)

[en] Radiation enhanced sublimation - the interstitial model to explain the radiation enhanced sublimation has been further proved. It predicts a threshold down to small energies similar as for the production of Frenhel pairs. A flux dependence is predicted, so that the yield cell decreases proportional to (P)³¹⁴ (P: Frenhel pair production rate). Chemical erosion of hydrocarbon films; which the chemical erosion of pure graphite differs significantly when using either thermal atomic hydrogen or energetic ions (yield increases from 5 x 10^-4 to ∼5 x 10^-2 C/H), a-C:H films are highly reactive against atoms alone. The erosion yield of a-C:H films when bombarded with thermal atoms is in the maximum 800k of the order 0/10^-1 and so similar as that of energetic ions on these films which is quite similar to the yield of pure graphite. It has been further found that hydrogen saturated graphite layers show also a high reactivity against atoms alone

[en] Recycling behavior in tokamaks with metals walls can be understood based on material parameters of diffusion constants, recombination rates, solubilities, etc. of hydrogen in metals. The expected recycling in fusion devices with carbon walls is believed to be determined by the saturation of graphite due to hydrogen impact. This would result in R 1 after a transient pumping with R < 1 and further change in the recycling would be only caused by temperature excursions of the walls. In contrast, present observations in tokamaks (JET, TFTR, TEXTOR) can not be explained by this simple picture. The final clarification of the observed phenomena is still open. The answer to this behavior will also drastically influence the expected tritium inventory in fusion machines

[en] This paper gives an overview of the research activities at TEXTOR on impurity production, impurity transport through the plasma, and then deposition. First, laboratory experiments on chemical erosion by hydrogen and oxygen and radiation-enhanced sublimation are described, followed by the main part, which concentrates on the TEXTOR data of impurity release, impurity transport, and redeposition. The differences between the behavior of high-Z and low-Z materials are discussed. Many of the TEXTOR experiments are carried out using special limiter locks, but the overall carbon balance of net erosion sources and net deposition zones are also shown. Finally, modeling of erosion and dedicated transport experiments are addressed

[en] The International Tokamak Experimental Reactor (ITER), the first burning fusion plasma experiment based on the tokamak principle, is ready for construction. It is based on many years of fusion research resulting in a robust design in most of the areas. Present day fusion research concentrates on the remaining critical issues which are, to a large extent, connected with processes of plasma-wall interaction. This is mainly due to extended duty cycle and the increase of the plasma stored energy in comparison with present-day machines. Critical topics are the lifetime of the plasma facing components (PFC) and the long-term tritium retention. These processes are controlled mainly by material erosion, both during steady state operation and transient power losses (disruptions and edge localized modes (ELMs)) and short- and long-range material migration and re-deposition. The extrapolation from present-day 'full carbon wall' devices suggests that the long-term tritium retention in a burning fusion device would be unacceptably high under these conditions allowing for only an unacceptable limited number of pulses in a D-T mixture. As a consequence of this, research activities have been strengthened to understand in more detail the underlying processes of material erosion and re-deposition, to develop methods to remove retained tritium from the PFCs and remote areas of a fusion device and to explore these processes and the plasma performance in more detail with metallic PFC, such as beryllium (Be) and tungsten (W), which are foreseen for the ITER experiment. This paper outlines the main physical mechanisms leading to material erosion, migration and re-deposition and the associated fuel retention. It addresses the experimental database in these areas and describes the further research strategies that will be needed to tackle critical issues

[en] Proper wall conditioning has been a major element in the development of fusion energy on the way to achieve high fusion plasma performance. Various of these techniques have been pioneered in the TEXTOR tokamak and later applied successfully in various devices worldwide. The main issues are to clean the surface from surface-bounded impurities, to remove hydrogen, and to coat the entire wall surface with a thin film of a proper first-wall material. The main benefits of wall conditioning are to control the oxygen impurity content of the plasma and to offer a suitable first-wall material. Entire coating of the first wall has allowed one to control to some extent the recycling hydrogenic fluxes but in particular to study the complex coupling between the choice of wall materials and the behavior of the plasma edge. This paper presents a review of the different wall-conditioning methods used in TEXTOR and their effects on the plasma behavior. Also, new wall-conditioning concepts, compatible with steady-state magnetic fields, are outlined briefly

[en] The present fusion experiences are largely based on devices with graphite as first wall material. The graphite choice allows for a large operational space in all plasma scenarios, with small detrimental effect of impurities on the main plasma performance (low Z) and for power loads in transients and steady state exceeding occasionally the material limits. However, the chemical affinity of graphite with hydrogen results in large carbon erosion, migration and redeposition, which would lead to unacceptable T retention in ITER, based on extrapolation from present devices. The ITER wall material choice is determined by aim to combine the advantages of C (no melting) at the high heat flux areas, of Be on the large scale first wall (low Z) and of W on the upper divertor baffle (low erosion, large lifetime) while minimising simultaneously the critical issues of C (T retention) and of W (plasma contamination). This material choice will be tested first in the new JET ITER like wall project which serves then to define the ITER operational conditions such that they are compatible with the wall requirements and scientific goals of ITER. The wall choice in ITER is still largely determined by the possibility for detrimental conditions with respect to transient power loads, the main plasma contamination with high Z impurities and also the relatively low duty cycle and moderate amount of neutrons. In contrast, for DEMO the development of plasma control must allow the use of first wall materials that are optimised fully by the materials requirements of mechanical stability, low T retention, low erosion (large lifetime) and low activation. This is the challenge in fusion both for plasma physics and control and material physics. (orig.)

[en] Low energy helium (100 eV-5 keV) has been implanted in single crystal nickel samples at 80 K and approx.= 35 K. Subsequently the helium release has been measured upon heating the sample up to 1400 K. A pronounced release peak at (55 +- 10) K has been observed from which an activation energy of (0.14 +- 0.03) eV can be deduced. The observed release process is ascribed to interstitial diffusion of helium in nickel. (orig.)

[en] A new system for routine digitization of video images has been implemented on TEXTOR-94. This video images system is used to analyse the impurity production, recycling and power deposition on limiters and to study the structure and location of visible emissions from the main plasma. The video system is based on a PC and is characterised by a high space and temporal resolution of video capture, storage and retrieval. On TEXTOR-94 cameras with CCIR norm take video pictures at 50 half frame per second with a resolution of 728 x 480 pixels. The video signals are captured from the cameras with a commercial framegrabber board and transferred with about 100Mbytes/sec directly into the main memory, which then processes every conceivable video signal in real time. The captured images are stored automatically on the hard disc of the PC. A special data display and analysis program has been developed to analyse the 2D images from spectroscopic, thermographic and other observations allowing an easy and comfortable handling of stored images. In this paper, we give a short description of the video system concentrating mostly on guidelines for the use of the VIDEO program. (orig.)