J. W. Coenen

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.

First scenario development with the JET new ITER-like wall

In the recent JET experimental campaigns with the new ITER-like wall (JET-ILW), major progress has been achieved in the characterization and operation of the H-mode regime in metallic environments: (i) plasma breakdown has been achieved at the first attempt and X-point L-mode operation recovered in a few days of operation; (ii) stationary and stable type-I ELMy H-modes with beta(N) similar to 1.4 have been achieved in low and high triangularity ITER-like shape plasmas and are showing that their operational domain at H = 1 is significantly reduced with the JET-ILW mainly because of the need to inject a large amount of gas (above 10(22) Ds(-1)) to control core radiation; (iii) in contrast, the hybrid H-mode scenario has reached an H factor of 1.2-1.3 at beta(N) of 3 for 2-3 s; and, (iv) in comparison to carbon equivalent discharges, total radiation is similar but the edge radiation is lower and Z(eff) of the order of 1.3-1.4. Strong core radiation peaking is observed in H-mode discharges at a low gas fuelling rate (i. e. below 0.5 x 10(22) Ds(-1)) and low ELM frequency (typically less than 10 Hz), even when the tungsten influx from the diverter is constant. High-Z impurity transport from the plasma edge to the core appears to be the dominant factor to explain these observations. This paper reviews the major physics and operational achievements and challenges that an ITER-like wall configuration has to face to produce stable plasma scenarios with maximized performance.

First operation with the JET International Thermonuclear Experimental Reactor-like wall

To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITERs plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixtu...

Development of tungsten fibre-reinforced tungsten composites towards their use in DEMO—potassium doped tungsten wire

For the next step fusion reactor the use of tungsten is inevitable to suppress erosion and allow operation at elevated temperature and high heat loads. Tungsten fibre-reinforced composites overcome the intrinsic brittleness of tungsten and its susceptibility to operation embrittlement and thus allow its use as a structural as well as an armour material. That this concept works in principle has been shown in recent years. In this contribution we present a development approach towards its use in a future fusion reactor. A multilayer approach is needed addressing all composite constituents and manufacturing steps. A huge potential lies in the optimization of the tungsten wire used as fibre. We discuss this aspect and present studies on potassium doped tungsten wire in detail. This wire, utilized in the illumination industry, could be a replacement for the so far used pure tungsten wire due to its superior high temperature properties. In tensile tests the wire showed high strength and ductility up to an annealing temperature of 2200 K. The results show that the use of doped tungsten wire could increase the allowed fabrication temperature and the overall working temperature of the composite itself.

Divertor tungsten tile melting and its effect on core plasma performance

For the 2007 and 2008 run campaigns, Alcator C-Mod operated with a full toroidal row of tungsten tiles in the high heat flux region of the outer divertor; tungsten levels in the core plasma were below measurement limits. An accidental creation of a tungsten leading edge in the 2009 campaign led to this study of a melting tungsten source: H-mode operation with strike point in the region of the melting tile was immediately impossible due to some fraction of tungsten droplets reaching the main plasma. Approximately 15 g of tungsten was lost from the tile over ~100 discharges. Less than 1% of the evaporated tungsten was found re-deposited on surfaces, the rest is assumed to have become dust. The strong discharge variability of the tungsten reaching the core implies that the melt layer topology is always varying. There is no evidence of healing of the surface with repeated melting. Forces on the melted tungsten tend to lead to prominences that extend further into the plasma. A discussion of the implications of melting a divertor tungsten monoblock on the ITER plasma is presented.

Rotation and radial electric field in the plasma edge with resonant magnetic perturbation at TEXTOR

In this paper the results of a systematic experimental assessment of the plasma edge rotation and radial electric field with application of resonant magnetic perturbation (RMP) are presented. The results are based on the radially resolved measurement of the poloidal (vpol) and toroidal (vtor) rotation. It is shown that the radial electric field Er can be deduced from the radial force balance when small amplitude resonant magnetic perturbations are applied to the plasma boundary (Br/Btor ~ 10?4).Both vpol and vtor spin-up in the ion-diamagnetic-drift and co-current direction, respectively, with increasing external perturbation field (?vpol ~ 15?km?s?1, ?vtor ~ 2?5?km?s?1) yielding an increase in Er by ?Er,max = 9?kV?m?1. The toroidal rotation increases over the whole radius while the poloidal rotation shows distinct local features driving the evolution of the Er-profiles. Depending on the edge safety factor a local (at the q = 5/2 rational surface) increase in the shear rate ?E?B (??q=5/2 = 1.4 ? 105?s?1) or reduced shearing can occur. Increased shearing is correlated with an improved particle confinement with an increase in the particle confinement time by ??p = +40%. Increasing the local resonant amplitude by 30% induces a reduced density level, the so-called RMP induced pump-out. At this confinement stage the shear rate decreases by 15% correlated with a significant drop in particle confinement (??p = ?30%).Field line tracing in the vacuum approximation gives indications towards explaining the threshold behaviour connecting the shearing rate, confinement stages and magnetic topology to the amount of applied RMP. However, this basic approach does not account for plasma response and the results presented are linked in the discussion section to recent results on the link between rotation and plasma response as well as on the transport features of RMP.

Beryllium Migration in JET ITER-like Wall Plasmas

JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (Ein = 35 eV) and more than 100%, caused by Be self-sputtering (Ein = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at Ein = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.

Materials for DEMO and reactor applications-boundary conditions and new concepts

DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER towards realizing fusion. For the realization of fusion energy especially, materials questions pose a significant challenge already today. Heat, particle and neutron loads are a significant problem to material lifetime when extrapolating to DEMO. For many of the issues faced, advanced materials solutions are under discussion or already under development. In particular, components such as the first wall and the divertor of the reactor can benefit from introducing new approaches such as composites or new alloys into the discussion. Cracking, oxidation as well as fuel management are driving issues when deciding for new materials. Here composites as well as strengthened CuCrZr components together with oxidation resilient tungsten alloys allow the step towards a fusion reactor. In addition, neutron induced effects such as transmutation, embrittlement and after-heat and activation are essential. Therefore, when designing a component an approach taking into account all aspects is required.

Influence of the dynamic ergodic divertor on transport properties in TEXTOR

Experiments to investigate transport properties under the influence of the dynamic ergodic divertor (DED) on TEXTOR are discussed. Relativistic runaway electrons are applied for studying transport properties of ergodization such as enhanced runaway loss. The ergodization causes an enhanced loss rate; this loss is higher for low relativistic electrons than for highly relativistic ones, in good agreement with particle orbit mapping. Edge transport can be controlled by the DED perturbation: in limiter H-mode plasmas ELM-like particle and heat bursts associated with the formation of enhanced edge pressure gradients are mitigated in the 6/2 configuration on the expense of a reduced pedestal height. Finally, the plasma is driven back to L-mode under the influence of the magnetic perturbation. In the 3/1 configuration the onset of tearing modes limits the possibility to affect edge transport. A mode of spontaneous density built-up has been found for the TEXTOR-DED as well. This mode is in particular strong for an inward shifted plasma; the built-up has a resonant character with respect to q(a). Langmuir probe measurements with two probe arrays show a strong influence of the magnetic ergodization on both the edge plasma equilibrium and fluctuation parameters. In particular, in the ergodic zone the turbulence properties and turbulence-driven flux are profoundly modified.