A. Kreter

Plasma?surface interaction, scrape-off layer and divertor physics: implications for ITER

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITERs use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Tritium retention in next step devices and the requirements for mitigation and removal techniques

Mechanisms underlying the retention of fuel species in tokamaks with carbon plasma-facing components are presented, together with estimates for the corresponding retention of tritium in ITER. The consequential requirement for new and improved schemes to reduce the tritium inventory is highlighted and the results of ongoing studies into a range of techniques are presented, together with estimates of the tritium removal rate in ITER in each case. Finally, an approach involving the integration of many tritium removal techniques into the ITER operational schedule is proposed as a means to extend the period of operations before major intervention is required.

Analysis of tungsten melt-layer motion and splashing under tokamak conditions at TEXTOR

Behaviour and characteristics of W plasma-facing components under impinging high heat fluxes are investigated in view of the material choices for the divertor in future devices such as ITER and DEMO. Experiments have been carried out in the plasma edge of the TEXTOR tokamak to study melt-layer motion, macroscopic tungsten erosion from the melt layer as well as the changes in material properties such as grain size and abundance of voids or bubbles. The parallel heat flux at the radial position of the plasma-facing components (PFCs) in the plasma ranges around q|| ~ 45?MW?m?2 allowing samples to be exposed at an impact angle of 35? to 20?30?MW?m?2. Melt-layer motion perpendicular to the magnetic field is observed following a Lorentz force originating from thermoelectric emission of the hot sample. Up to 3?g of molten W are redistributed forming mountain-like structures at the edge of the sample. The typical melt-layer thickness is 1?1.5?mm. Those hills are, due to the changes in the local geometry, particularly susceptible to even higher heat fluxes of up to the full q||. Locally the temperature can reach up to 6000?K, high levels of evaporation are causing significant erosion in the form of continuous fine-spray (~1 ? 1024?atoms?m?2?s?1). Strong evaporation cooling is observed hindering the further heating of the samples. In addition, the formation of ligaments and splashes occurs several times during the melt phase ejecting droplets in the order of several 10??m up to 100??m probably caused by an instability evolving in the melt. In terms of material degradation several aspects are considered: formation of leading edges by redistributed melt, bubble formation and recrystallization. Bubbles are occurring in sizes between 1 and 200??m while recrystallization increases the grain size up to 1.5?mm. The power-handling capabilities are thus severely degraded. Melting of tungsten (W) in future devices is highly unfavourable and needs to be avoided especially in light of uncontrolled transients and possible unshaped PFCs

Investigation of the impact of transient heat loads applied by laser irradiation on ITER-grade tungsten

Cracking thresholds and crack patterns in tungsten targets after repetitive ITER-like edge localized mode (ELM) pulses have been studied in recent simulation experiments by laser irradiation. The tungsten specimens were tested under selected conditions to quantify the thermal shock response. A Nd:YAG laser capable of delivering up to 32 J of energy per pulse with a duration of 1 ms at the fundamental wavelength λ = 1064 nm has been used to irradiate ITER-grade tungsten samples with repetitive heat loads. The laser exposures were performed for targets at room temperature (RT) as well as for targets preheated to 400 °C to measure the effects of the ELM-like loading conditions on the formation and development of cracks. The magnitude of the heat loads was 0.19, 0.38, 0.76 and 0.90 MJ m−2 (below the melting threshold) with a pulse duration of 1 ms. The tungsten surface was analysed after 100 and 1000 laser pulses to investigate the influence of material modification by plasma exposures on the cracking threshold. The observed damage threshold for ITER-grade W lies between 0.38 and 0.76 GW m−2. Continued cycling up to 1000 pulses at RT results in enhanced erosion of crack edges and crack edge melting. At the base temperature of 400 °C, the formation of cracks is suppressed.

Experience with bulk tungsten test-limiters under high heat loads: melting and melt layer propagation

The paper provides an overview of processes and underlying physics governing tungsten melt erosion in the fusion plasma environment. Experiments with three different bulk tungsten test-limiters were performed in TEXTOR: (i) thermally insulated solid plate fixed on a graphite roof-like limiter heated up by the plasma to the melting point, (ii) macro-brush of the ITER-relevant castellated structure and (iii) lamellae structure developed for the JET divertor. The main objectives were to determine the metal surface damage, the formation of the melt layer and its motion in the magnetic field. PHEMOBRID-3D and MEMOS-1.5D numerical codes were used to simulate the experiment with the roof-like test-limiter. Both experiments and simulation showed that the melting of tungsten can lead to a large material redistribution due to thermo-electron emission currents without ejection of molten material to the plasma.

Radio frequency ion source for plasma diagnostics in magnetic fusion experiments

Low-divergent quasistationary neutral beams are often applied in modern magnetic fusion devices as a diagnostic tool providing unique information about plasma parameters. The most important requirements of these beams are sufficiently large current and energy of the particles, so that the beam can penetrate to the plasma core. Also the duration of the beams must be long enough, i.e., close to that of a plasma discharge, amounting to at least a few seconds for large fusion devices. We developed a neutral beam injector for plasma diagnostics in the tokamak TEXTOR-94 which is capable of meeting these requirements. The maximum beam energy is 50 keV and the source operated in hydrogen delivers an ion current of up to 2 A with a pulse duration of up to 4 s. The low divergent beam (∼0.5°– 0.6°) is geometrically focused 4 m downstream from the source having a 1/e width of ∼ 70 mm at the focal point. The beam can be modulated with a frequency variable up to 500 Hz. The ion source plasma is produced by a radio freq...

Linear Plasma Device PSI-2 for Plasma-Material Interaction Studies

Abstract The linear plasma device PSI-2 serves as a pilot experiment for the development of components, operational regimes and control systems for the linear plasma device JULE-PSI, which will be located in the nuclear environment allowing studies of radioactive and toxic samples. PSI-2 is also used for fusion reactor relevant plasma-material interaction studies. This contribution describes the PSI-2 layout and parameters and summarizes the recent scientific and technical progress in the project, including the installation of a target station for the sample manipulation and analyses.

Effect of surface roughness and substrate material on carbon deposition in tokamak TEXTOR

The technique of 13 CH4 tracer injection through test limiters was applied to study the influence of surface roughness and substrate material on the local 13 C deposition. Spherically shaped graphite limiters with integrated gas injection were prepared with two different grades of surface roughness, ~0.1 μm and ~1 μm. In addition, tungsten limiters with a roughness of ~0.1 μm were used to study the substrate material effect. The limiters were exposed to the SOL plasma for two discharge scenarios, Ohmic and neutral beam heated. The 13 C deposition efficiency - the ratio of the locally deposited to the injected amount of 13 C - and the deposition pattern were evaluated by post-mortem surface analysis. Surface roughness has a pronounced effect and increases the 13 C deposition efficiency on graphite with rougher surface by a factor of 3-5 compared to smoother graphite. On tungsten a factor of 2-4 less carbon is deposited than on graphite with similar surface roughness. A systematically higher amount of 13 C (by a factor 1.5-2.5) was deposited on limiters exposed to Ohmic compared to neutral beam heated plasmas.

Active control of type-I edge localized modes on JET

The operational domain for active control of type-I edge localized modes (ELMs) with an n = 1 external magnetic perturbation field induced by the ex-vessel error field correction coils on JET has been developed towards more ITER-relevant regimes with high plasma triangularity, up to 0.45, high normalized beta, up to 3.0, plasma current up to 2.0 MA and q95 varied between 3.0 and 4.8. The results of ELM mitigation in high triangularity plasmas show that the frequency of type-I ELMs increased by a factor of 4 during the application of the n = 1 fields, while the energy loss per ELM, ΔW/W, decreased from 6% to below the noise level of the diamagnetic measurement (<2%). No reduction of confinement quality (H98Y) during the ELM mitigation phase has been observed. The minimum n = 1 perturbation field amplitude above which the ELMs were mitigated increased with a lower q95 but always remained below the n = 1 locked mode threshold. The first results of ELM mitigation with n = 2 magnetic perturbations on JET demonstrate that the frequency of ELMs increased from 10 to 35 Hz and a wide operational window of q95 from 4.5 to 3.1 has been found.

Capture by aerogel—characterization of mobile dust in tokamak scrape-off layer plasmas

The aim of this letter is to demonstrate the feasibility and potential of the novel in situ dust diagnostic method-capture by aerogel targets. Aerogel, a highly porous material with a density of a ...