[en] Processes of mixed-material surface evolution and corresponding material properties are investigated at several experiments at IPP Garching. The following topics will be presented and discussed: 1. Studies on co-bombardment of tungsten by D and C. Co-bombardment of tungsten by D and C leads to synergistic effects, which result in two principal dynamic regimes. One regime is characterised by continuous erosion of the tungsten surface while in the second regime one observes continuous growth of a carbon layer on the tungsten substrate. The processes involved and their influence on the boundary between these regimes will be discussed. 2. Surface processes in the formation of layers with binary and ternary ITER relevant material combinations (Be/C/W). Migration and redeposition of wall materials in fusion machines leads to formation of mixed material surface layers. At elevated temperatures carbides and/or metal alloys will form, which have a significant influence on erosion properties and fuel isotope retention. The principal processes have been identified and discussed. 3. Formation of mixed-material layers leads to significant modifications of fuel retention and fuel isotope permeation properties of the substrate materials. Results of corresponding studies for the ITER-relevant material combinations will be presented and discussed. (author)

[en] With several military base closures resulting in property transfer to public use and the decommissioning of many legacy waste facilities, the opportunity for remediation of older buildings is increasing. Along with these projects, come several problems that could give the potential remediator some surprises. During the preconstruction and planning phases of the original construction activities, several generations of drawings were most likely produced for approval and permit submittal. Over the years, buildings may undergo several renovations with or without the full characterization or remediation that should be done when radioactive materials are used on a site. New walls or floors may be built over the original construction materials. Contamination in and around the building may have resulted from processes that were accepted at the time or from inadvertent activities that may have been covered up, including accidental spills. Many buildings contain hidden rooms or accesses that over time became useless and have been closed up or over, these areas may not be very obvious. When characterizing a building the effluents of the building are usually forgotten, sewer lines are important areas to investigate. All these items could cause a remediator to overlook a potentially highly contaminated area. With more of these facilities being turned over for public use, correctly characterizing these buildings will become a more common problem

[en] The simultaneous bombardment of a W surface by hydrogen isotopes and carbon impurities leads to synergistic effects with significantly different plasma-wall interaction properties compared to the ones of pure hydrogen or pure carbon bombardment. Simultaneous impact of carbon and hydrogen ions has been realised by irradiation of W targets with CH₃⁺ radicals produced by a high current ion source. For comparison, the same experiment was also carried out with C⁺ ions. The measured evolution of the target weight as a function of irradiation fluence and the final amount of deposited carbon are compared to a simple model of erosion-deposition balance. This allows one to derive the dependence of the carbon sputtering yield on both the impacting ion species and the target temperature

[en] The impurity concentrations and corresponding Z_eff contributions as well as the dilution of the deuterium background plasma in ASDEX are determined by VUV spectroscopy. The methods used are described in detail. We describe the absolute calibration of our VUV survey spectrometer with two different calibration sources, as well as our ZEDIFF time-dependent transport code, used for interpreting the spectroscopic measurements. The assessed spectroscopic Z_eff compares quite well with the bremsstrahlung Z_eff as demonstrated for a number of representative ohmically and additionally heated discharges. In order to obtain these results readily on a shot-to-shot basis at the end of each discharge, a simplified fast evaluation method is introduced. This fast analysis method yields the central impurity concentrations, the central Z_eff contributions, and the dilution of the deuterons. Again, the results from the fast analysis method agree well with those from our extended transport code treatment and with the bremsstrahlung Z_eff. (orig.)

[en] Impurity transport in tokamaks has been investigated for several reasons. On the one hand, there is the possibility of attaining intolerably large radiation losses and deuterium/tritium dilution in the plasma core region caused by impurities. On the other hand, the behaviour of impurity ions can give valuable information on the physical nature of the underlying transport mechanisms, which is of great interest from a theoretical point of view. Therefore it is desirable to measure as accurately as possible the impurity transport coefficients for various ions under different plasma conditions. In the following we describe a new experimental method - hitherto used to analyze the transport of the background plasma - that was recently applied to impurities on ASDEX. The principles of the method are outlined with particular emphasis on the differences that occur in applying the technique to impurity ions instead of the bulk plasma. Experimental results are shown from ASDEX discharges with pure ohmic and with additional NI-heating derived from analysis with a simplified transport model mainly used for less detailed study of the transport quantities as a function of plasma parameters. (author) 6 refs., 1 tab

[en] Pellet injection into ASDEX discharges allows considerable improvement of the confinement properties. Simultaneously with this improvement a strong accumulation of metallic impurities is observed, which leads to intolerable cooling of the plasma core region. We discuss the experimental phenomena and the underlying transport changes associated with the accumulation. (orig.)

[en] A critical issue for the choice of main chamber first wall materials in future fusion devices such as ITER is the erosion rate due to bombardment by charge-exchange (CX) neutrals. Due to the relatively small flux density of impacting particles, respective measurements are only possible using long term samples (LTS) exposed for a full experimental campaign. In ASDEX Upgrade, CX erosion has been studied extensively for tungsten on the inner heat shield by placing four W coated tiles at different poloidal positions in one toroidal sector. During the same campaign, several LTS were placed at different poloidal and toroidal positions of the outer wall. ¹³C and Cu coated graphite probes were also used in order to test and compare W low and medium Z alternatives. The erosion results from the probes are compared with the calculated erosion [W. Eckstein, C. Garcia-Rosales, J. Roth, W. Ottenberger IPP Report, IPP 9/82]; [H. Verbeek, J. Stober, D. Coster, W. Eckstein, R. Schneider Nucl. Fus. 38 (1998) 12] and a figure of merit (F. of M.) between several materials is proposed which also takes into account the plasma isotope effect in CX erosion

[en] A quick and simple experimental method of comparing the global retention capabilities of different divertor configurations for gaseous impurities is presented. Short gas puffs into the main chamber and analysis of the time dependence of central impurity lines are used to determine the retention capability in terms of time constants, which characterize the impurity leakage rate from the divertor chamber to the main plasma and to the pumping system. A simple two-reservoir model is used to define these basic parameters of divertor retention. For a more detailed evaluation of the divertor retention parameters from spectroscopic data, an impurity transport code is used to simulate the impurity radiation measured in various experiments. The dependence of the retention capability on the divertor plasma density and temperature as well as on the distance along the field lines back to the main plasma is discussed. The dependence on the impurity species is also investigated. On the basis of these results, the consequences of the divertor retention for helium exhaust in future fusion devices are analysed by numerical simulation. (author). 32 refs, 7 figs, 4 tabs

[en] One of the main disadvantages of carbon as first wall material in a fusion device is the co-deposition of hydrogen with the eroded carbon. These layers will contain a significant amount of the tritium inventory of a fusion reactor. After venting brownish layers and flakes were found under the divertor structure of ASDEX Upgrade. First investigations were made on these flakes. Due to the complicated structure beyond the divertor, shadowing effects occur indicating that the brown layers are deposited by ionised particles. The flakes were analysed using SEM and ion beam techniques. Two different types of hydrocarbon layers were found: The brownish hydrogen poor layer (D/C = 0.4) and transparent hydrogen rich layer (D/C = 1) The total amount of carbon beyond the divertor could be estimated to 1.5 g, deposited in 3000 s of plasma discharges. First measurement of the layer growth using a quartz crystal microbalance instrument is presented

[en] In fusion devices with magnetic plasma confinement, plasma facing vessel surfaces are exposed to a continuous flux of particles and energy from the plasma. These lead to various erosion processes with corresponding release of wall material as impurities into the plasma. Due to transport processes, both in the plasma boundary and in the confined plasma region, eroded wall material migrates to other vessel locations were it is to varying extent deposited and re-eroded. The continuous material redistribution has significant implications for wall component lifetime, wall integrity and radiation safety. After an introduction to the fundamental transport processes, the lecture will focus on the experimental methods for the investigation of material transport and migration processes. In addition to the respective methodology, corresponding key observations in present day fusion devices and their consequences for design and operation of ITER will be discussed. This document is made of the slides of the presentation. (author)