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In fusion devices, the retention of the fusion fuel deuterium (D) and tritium (T) in plasma-facing components (PFCs) is a major concern. Measurement of their hydrogen isotope content gives insight into the retention physics. In FREDIS, two methods of thermal desorption are used for retention measurements: In Thermal Desorption Spectrometry (TDS) the samples are heated by 6 infrared lamps up to 1433 K with linear temperature ramps of up to 1.67 K/s. The desorbed gases are detected up to 100 amu/e with a double-QMS (Quadrupole Mass Spectrometer) that can distinguish between helium and D2 and uses an innovative differential pumping system. In a connected vacuum chamber, a ∅3 mm spot can be heated on the sample surface by a high energy Nd:YAG laser pulse (E0

All optical and laser diagnostics in ITER will use mirrors to observe the plasma radiation. In the ITER environment, mirrors may become contaminated with plasma impurities hampering the performance of corresponding diagnostics. An in-situ mirror cleaning is proposed, which relies on ion sputtering of the contaminants and the affected mirror material. Previous research demonstrated the advantages of single crystal (SC) molybdenum (Mo) under sputtering conditions over polycrystalline concepts. Recently, the first single crystal rhodium (Rh) mirrors became available and tests have been started at the Forschungszentrum Julich. In a direct test, SC Rh and SC Mo mirrors were exposed under identical conditions in steady-state helium plasmas in the linear plasma device PSI 2. The energy of impinging ions was ∼100 eV matching conditions expected in the in-situ cleaning system in ITER. During exposure, molybdenum mirrors lost 420–500 nm due to sputtering. Rhodium mirrors lost about 1 μm. Exposure corresponded to 50–100 cleaning cycles in ITER. Nevertheless, rhodium mirrors have preserved their specular reflectivity, showing the maximum degradation of less than 7% at 250 nm. Molybdenum mirrors demonstrated a moderate decrease of specular reflectivity of 12–25%. Results open new perspectives for the use of single crystals in ITER diagnostics.

This paper focuses on encouraging results obtained on the characterization of RF produced plasmas during pulsed-mode wall conditioning discharges in ion cyclotron resonance frequency (ICRF) regime in the limiter tokamak TEXTOR. Recent Ion Cyclotron Wall Conditioning (ICWC) experiment carried out in TEXTOR tokamak, lead to the identification of various dependences of the antenna-plasma coupling efficiency on the plasma parameters for possible ICWC-discharge cleaning in ITER at half field. Our ICWC experiments emphasize on (i) study of antenna coupling during the mode conversion scenario, (ii) reproducible generation of ICRF plasmas for wall conditioning, by coupling RF power from one or two ICRF antennas and (iii) effect of application of an additional (along with toroidal magnetic field) stationary vertical (BV ≪ BT) or oscillating poloidal magnetic field (Bp ≪ BT) on antenna coupling and relevant plasma parameters.

New linear plasma devices are currently being constructed or planned in the Trilateral Euregio Cluster (TEC) to meet the challenges with respect to plasma surface interactions in DEMO and ITER: i) MAGNUM-PSI (FOM), a high particle and power flux device with super-conducting magnetic field coils which will reach ITER-like divertor conditions at high magnetic field, ii) the newly proposed linear plasma device JULE-PSI (FZJ), which will allow to expose toxic and neutron activated target samples to ITER-like fluences and ion energies including in vacuo analysis of neutron activated samples, and iii) the plasmatron VISION I. a compact plasma device which will be operated inside the tritium lab at SCK-CEN Mol, capable to investigate tritium plasmas and moderately activated wall materials. This contribution shows the capabilities of the new devices and their forerunner experiments (Pilot-PSI at FOM and PSI-2 Julich at FZJ) in view of the main objectives of the new TEC program on plasma surface interactions. (C) 2011 Forschungszentrum Julich, Institut fur Energieforschung-Plasmaphysik. Published by Elsevier B.V. All rights reserved.

In TEXTOR a high Z limiter experiment was carried out. Mo and W were used as a limiter (10 cm long, 6 cm wide and 5 cm high) and inserted up to 3 cm closer to the plasma than to the main limiter, an ALT-II graphite limiter. Both the thermal response of the limiter and the influence to the plasma performance were analyzed in detail and compared with those for the fine-grained graphite (EK98) routinely used. Maximum power fluxes of about 20 MW m −2 for 2 s could be loaded on the Mo limiter without severe damage, and the maximum surface temperature was about 2600 °C. Consequently the limiter surface was subjected to local melting leaving the trace of molten layers. However, the evaporated Mo during the melting did not influence the main plasma very much, probably owing to redeposition of Mo within one period of gyromotion after the ionization. After the experiment the surface of the Mo limiter was examined by various techniques. Although the limiter head was not seriously damaged, appreciable recrystallization and grain growth were observed, which enhanced the propagation of intergranular cracking owing to the thermal stress triggered by congruent melting with thermocouple. No deposition of impurities of B, C, O and Si which were the main impurities in TEXTOR plasma, was observed on the limiter head. The very high temperature of the limiter surface seems to promote the evaporation of once-deposited impurities.

Cracking thresholds and crack patterns in tungsten targets have been studied in recent experiments after repetitive ITER-like ELM heat pulses in combination with plasma exposure in PSI-2 (Γtarget = 2.5–4.0 × 1021 m−2 s−1, ion energy on surface Eion = 60 eV, Te ≈ 10 eV). The heat pulses were simulated by laser irradiation. A Nd:YAG laser with energy per pulse of up to 32 J and a duration of 1 ms at the fundamental wavelength (λ = 1064 nm, repetition rate 0.5 Hz) was used to irradiate ITER-grade W samples with repetitive heat loads. In contrast to pure thermal exposure with a laser beam where the damage threshold under pure heat loads for ITER-grade W lies between 0.38 and 0.76 GW/m2, the experiments with pre-loaded W-samples as well as under combined loading conditions show a lower damage threshold of 0.3 GW/m2. This is probably due to deuterium embrittlement and/or a higher defect concentration in a region close to the surface due to supersaturation with deuterium. A pronounced increase in the D retention (more than a factor of five) has been observed during the combined transient heat loads and plasma exposure. Enhanced blister formation has been observed under these combined loading conditions.