C. Scientific and technological Programme
C.2. Plasma-Wall Interaction
mÜáäáééÉ=jÉêíÉåë=EfmmF ph.mertens@fz-juelich.de
Introduction
ITER is on the verge of being built and some of the physics questions related to plasma-wall inter-actions need urgent investigation in order to make the best engineering choices. JET will, for in-stance, study relevant material combinations in the frame of the ITER-like wall project. The main group on Plasma-Wall Interaction (PWI), which is traditionally organised in a topic-oriented fash-ion, has aligned in the course of year 2005 with the goals of the European Task-Force on PWI, as summarised on the website http://www.efda-taskforce-pwi.org/. The central motivation is the de-velopment of a viable solution to the questions of the ITER first wall and divertor materials. This policy ensures that ITER-relevant subjects are treated with due priority.
Improvement of basic understanding of plasma-surface interaction and related plasma processes near plasma-facing components are obviously part and parcel of the studies which were carried out.
It will be clear that the measurement and prediction of erosion and fuel retention on the one side, and the qualification of high-Z materials in a form suitable as plasma-facing components on the other side still are the critical questions currently investigated. Research in the Main Topic Group on Plasma-Wall Interaction thus deals with the following main fields:
• Erosion and deposition, carbon migration and tritium retention; these are the questions concerning possible ITER components conceived on a carbon basis. The development of re-moval techniques for hydrogen isotopes also belongs to this field, hereafter referred to as (1).
• High-Z plasma-facing components and mixed materials, for the case of an extended tungsten covering in ITER, possibly including the divertor plates. This is topic (2) below.
The Plasma-Wall Interaction group can rely on a large span of fusion facilities, i.e. beside TEXTOR also JET, AUG (ASDEX-Upgrade), DIII-D, PISCES, MAGNUM-Pilot, and others. Any specific physical investigation is thus carried out on the best suited machine. The full research programme is organised within the Trilateral Euregio Cluster, which encompasses the Belgian (ERM/KMS) and Dutch (FOM) partners in addition to the German Institutes in Jülich. The IEA partners (Japan, USA, Canada) are closely linked to this research programme as well, as can be seen in the following.
TEXTOR thus serves as the central fusion facility for the TEC partners, without prejudice to resort-ing to any other device upon demand.
1. Erosion and deposition, carbon migration and tritium retention
1.a. Chemical erosion and spectroscopy of hydrocarbons
The determination of chemical erosion yields of graphite is obviously required to realise how the erosion processes of carbon-based facing materials takes place. It is usually based on the detection of hydrocarbon break-up products such as C2 or CH by spectroscopy. The actual quantification is made with injection of a well-known amount of hydrocarbon gases i.e. CD4, C2D4, C2D6 etc. The ratio of particles injected into the plasma which are later on dissociated or ionised and the amount of light of the break-up products at the end of the chain, i.e. the C2 Swan band and the CD Gerö band, provide effective efficiency factors. The determination of these efficiency factors, their appli-cation on intrinsic hydrocarbon fluxes and thus the quantifiappli-cation of the erosion yield is topic of the ITPA D-SOL 2 task and the SEWG on chemical erosion within the EU Task Force on PWI. Ex-periments under the lead of a scientific coordinator from the Forschungszentrum have been per-formed at JET, TEXTOR, and ASDEX-Upgrade.
At TEXTOR, a series of experiments have been carried out to determine the number of photons of different break-up products per injected hydrocarbon molecule and to verify the present data bases on molecular and atomic data used in the ERO code. These experiments where performed with gas injection modules made of metal to exclude any surface effects. Different deuterated hydrocarbon gases have been injected into hydrogen plasmas to obtain not only the carbon-containing fragments but also to determine the amount of deuterium atoms produced per injected molecules. For typical ohmic TEXTOR plasmas with electron temperatures of about 50 eV at the LCFS we deduce the fol-lowing branching ratios:
CD Normalised on the number of injected carbon, we obtain the following ratio of CD light:
1.00 : 0.65 : 0.79 for CD4 : C2D4 : C2D6
C2 Normalised on the number of injected carbon, we obtain the following ratio of C2 light:
0.10 : 1.00 : 0.90 for CD4 : C2D4 : C2D6
Dβ Normalised on the number of injected molecules, we obtain the following ratio of Dβ light:
0.57 : 0.75 : 1.00 for CD4 : C2D4 : C2D6
Other transitions and species – a e.g. C or CD+ – in the range between 250 nm and 1000 nm have been simultaneously observed. All spectral observations together provide a so-called “footprint” of the initial hydrocarbon. The analysis of the data as well as accompanying ERO calculations for these experiments is ongoing.
Apart from the total amount of fragments produced during the dissociation or ionisation of the hy-drocarbon, we also investigate the internal structure (ro-vibrational population) of the molecular fragments at the end of the chain. For comparable plasma parameters no differences in the internal structure of e.g. C2 has been observed. Moreover, even sublimated C2 shows the same population as depicted in fig. 1. Thus, the information about the initial hydrocarbon species is lost during the dis-sociation process. The ro-vibrational population is determined by the local plasma conditions where the spectroscopic observed species is finally excited.
Fig. 1: Identical population of the Swan band for different initial hydrocarbon species.
At ASDEX-Upgrade, experiments have been performed to determine the branching ratio for the production of CH from C2H4 and CH4, and thus, to determine the role of higher hydrocarbons to the total erosion yield. The contribution of CH from C2H4 has been identified to be significant. The de-tailed analysis is ongoing. Additionally, the focus of the experiment was set to detached plasmas in the divertor. Light of the break-up products during hydrocarbon injection was observed, whereas the intrinsic signal was partially below the detection limit of the spectroscopic systems. Indications for a low chemical erosion under detached plasma conditions are given. Detailed analysis is in pro-gress.
1.b. Tritium retention: recycling of hydrogen isotopes
In order to understand the possible retention of tritium in a fusion device, detailed knowledge of the particle fluxes is of utmost importance. Fulcher-band spectroscopy is an example of a technique that has already been applied successfully in TEXTOR (see reports 2002–2004), ASDEX-Upgrade (AUG) and Tore Supra to determine the contribution of molecular hydrogen in the recycling hydro-gen flux in front of plasma-facing surfaces. This has been extended to JET where the properties of deuterium molecules and their contribution to the total deuteron flux in the outer divertor in L- and H-mode shots have been determined. This time it was tried to measure the H and T-fluxes in D-discharges via the observation of the isotopomers HD and TD. This method provides a very sensi-tive way to detect a low minority concentration (H or T) versus a high deuterium background level (see fig. 2).
Fig.2: Molecular flux ratio versus surface concentration.
Simultaneously, it offers a unique possibility to record and study molecular spectra of both tritium and tritiated hydrogen isotopes in a fusion boundary plasma environment, which had never been performed in the past (fig. 3).
Q-branch
-1.E+02 0.E+00 1.E+02 2.E+02 3.E+02 4.E+02 5.E+02 6.E+02 7.E+02 8.E+02
5990 6040 6090 6140 6190 6240 6290
wavelength / A
0-0 1-1 2-2 3-3 4-4
12 3 4 5 12345 12345 12345 12345
T
2Fig.3: Plasma boundary spectra of T2.
Another study was started with the aim to assess the potential of Doppler spectral line shapes as a diagnostic of the statistical properties of turbulence in edge plasmas of tokamaks. This technique would rely on a careful analysis of time-integrated Doppler spectra emitted by charge exchange neutrals, the tails of which are sensitive to non-Gaussian turbulent fluctuations. In particular, a strong effort was devoted to the modelling of the contribution of the neutrals created by charge ex-change in the plasma core, using EIRENE. The latter have indeed to be taken into account when modelling the high energy part of the neutrals velocity distribution. More information can be found in the report of the theory group.
Besides that, a comparison of the hydrogen flux determined by two different methods has been per-formed: Hα emission based versus Langmuir probe flux measurements. The plasma flux onto the bumper limiter (i.e. the DED tiles) was measured by means of an Hα flux diagnostic with a correc-tion regarding the local electron density estimated from the measurement of Hα/Hγ line intensity ra-tios. This flux was compared with the flux measured by a Langmuir probe installed in a DED tile.
Discharges with a density ramp and with the plasma shifted towards the bumper limiter were used for this comparison. There is a difficulty for a direct comparison because the Hα based diagnostic provides the plasma flux perpendicular to the surface while the Langmuir probe provides this quan-tity parallel to the magnetic field lines. For this reason the angle between the magnetic field lines and the bumper limiter surface was used as a fitting parameter. The results of the measurements are shown in fig. 4. The electron density measured with the help of the Hα/Hγ line intensity ratio is higher because the Hα and Hγ emissions originate from the same but finite distance above the sur-face while the Langmuir probe measures plasma parameters close to the bumper limiter sursur-face. As a result, the plasma flux measured by means of the Hα diagnostic very well coincides with the Lang-muir probe data for an angle of about 1.1°.
0 1 2 3 4
parralel flux / 1019 cm-2 s-1
time / s
Hα data (1.1° angle) Langmuir probe
Fig. 4: Electron densities and plasma fluxes measured by the Hα based diagnostic and by a Langmuir probe at the bumper limiter.
1.c. Fuel accumulation, carbon migration and materials mixing in gaps of castellated structures
Investigations have continued addressing the molybdenum limiter with ITER-like castellation being exposed to the scrape-off layer of TEXTOR under erosion-dominated conditions.
The extensive set of surface diagnostics was used to study the deposited layers and deuterium fuel accumulation in the gaps of a castellated limiter: namely Secondary Ion Mass-Spectrometry (SIMS), Nuclear Reaction Analysis (NRA), Electron Probe Micro-Analysis (EPMA), depth profil-ing with a Dektak stylus profiler and colourimetry. Top surfaces of a castellation were found to be clean from deposited layers and deposits were found in gaps only. A local erosion zone has devel-oped on the plasma-exposed sides of gaps. The thickness of deposited layers decays exponentially with the depth of gaps showing a characteristic e-folding length of 1.2 to 1.7 mm. These findings correlate with the results from the other machines which show the common nature of deposition processes in gaps.
Massive metal intermixing was detected in the gaps with a molybdenum content of at least 65 at.-% in the deposit. The molybdenum originates from local erosion on the plasma open side of the gap. The amount of fuel accumulated in the gaps was estimated to be 0.02 % to 0.04 % of the impinging plasma fluence.
An attempt was made to model the deposition patterns in the gaps. Good agreement of modelling results with experimental data was achieved assuming a particle reflection coefficient of 0.5 for carbon atoms trapped in the gaps. This outlines the importance of reflection while assessing the processes in gaps.
However, more experimental data on the edges of gaps are needed to allow for a better comparison with modelling. These investigations were made in the framework of the multi-machine ITPA (Task DSOL-13) of the ITPA TG on Divertor Physics and SOL.
Fig. 5: Castellated limiter after exposure in the SOL plasma of TEXTOR.
Comparison of the modelling results with experimental data: 1) Modelling results, 2) Normalized data from SIMS investigations, 3) Normalized data from EPMA measurements.
1.d. Removal of carbon layers by oxygen cleaning
In case tritium is co-deposited with carbon onto plasma-facing surfaces, techniques have to be de-veloped to remove the thin amorphous layer which contains hydrogen isotopes. Glow Discharge Cleaning (GDC) is one of them. Removal of re-deposited C layers by oxygen GDC has been ex-plored in TEXTOR in an RF supported (13.2 MHz, ∼ 250 W) DC glow discharge in gas mixtures of O2/He at different ratios. The carbon removal rate was evaluated from the CO and CO2 partial pres-sures and the pumping speed. It amounts to about 2.1 x 1019C/s, summing up to about 5.22 g of C removed from the vessel during the GDC treatment. The addition of He to O with at a constant O pressure did not influence the production of CO and CO2. Fig. 6 shows the removal of an amor-phous carbon layer (a-C:H) of about 190 nm that had been deposited beforehand on silicon probes.
Fig 6: Appearance of Si probes coated with an a-C:H film before and after O2 GDC treatment.
1.e. ICRF Discharge conditioning (helium-oxygen mixtures for removal of hydrogen isotopes)
ICRF Discharge Conditioning (ICRF-DC) is proposed as an alternative technique for wall condi-tioning between shots in the next generation of super-conducting fusion machines like ITER or Wendelstein 7-X. The point is that the main magnetic field of these devices cannot be switched off for such short periods and the GDC method discussed above cannot be applied as long as the field is present. The inter-machine studies of the topic have been performed on the present-day circular (TEXTOR) and divertor-type (ASDEX-Upgrade and JET) tokamaks.
Recent activities on TEXTOR have focussed on the study of removal of carbon layers by oxygen treatment. Long term retention of tritium fuel in the surface or bulk material of plasma facing com-ponents is one of the major problems for a fusion reactor. Oxidation of the re-deposited carbon lay-ers is one of the most promising tools to remove fuel from the walls. In order to prove the feasibility of the proposed cleaning technique in the presence of permanent intense magnetic fields, the ICRF-DC in a He/O2 gas mixture has successfully been tested for the first time and compared with the standard glow discharge conditioning (GDC). ICRF discharges of 5 sec duration ignited in He and with oxygen added (about 18 mbar ) during the RF pulse showed a complete consumption of the molecular oxygen and a transformation to CO and CO
A
2. However, the carbon oxides were mainly released after the RF pulses, while the removal during the pulse was limited by the upper pressure limit for the ICRF antenna ensuring a non-arcing operation. Assuming a duty cycle of 1:10, the overall C-removal rate would be similar to that observed in GDC. Increasing the C-removal rate would require higher pumping speeds and, probably, a higher oxygen injection rate in between RF pulses. This will be a subject for further study.
Fig. 7: Temporal behaviour of CO, CO2 ,O2 and HD partial pressures during and after the ICRF pulse.
The ICRF pulse lasted from 1.0 s to 3.9 s and oxygen was injected at 1.1 s.
In the same area of wall-cleaning by RF-produced plasmas, experiments in collaboration have been performed on AUG and JET aiming at optimising ICRF-DC in divertor-type machines. A new rec-ipe for safe and reliable RF plasma production with improved antenna coupling and improved ra-dial/poloidal homogeneity was proposed and successfully tested on both machines. First compari-sons of ICRF-DC versus GDC in terms of the argon removal rate using walls preloaded with argon were made in ASDEX-Upgrade.
1.f. Modelling of erosion and re-deposition with the ERO code and associated experiments
The modelling of erosion and re-deposition is expected to deliver predictive information on the be-haviour of different materials, especially carbon, in the geometry and environmental conditions of the coming fusion devices. That is the reason why it is so important, and why accompanying experi-mental effort has been invested to provide the code calculations with sufficient interpretable data for comparison, as will be seen in the following.
Experimental work
13CH4 tracer injection is a powerful tool to investigate the local carbon transport. Well-defined amounts of 13CH4 are injected through a hole in a test limiter into the running discharge. Camera in-spection allows us to study the evolution of the deposition pattern during each plasma pulse; local spectroscopy observes spectral light distribution during injection. Post-mortem analysis can distin-guish between injected 13C and 12C from the background plasma, hence the 13C deposition effi-ciency (the number of locally deposited 13C atoms divided by the number of injected 13C atoms) can be determined.
Experiments were performed to investigate the influence of the substrate material and limiter ge-ometry on the 13C transport. Roof-like test limiters were covered by plane plates of polished graph-ite, tungsten and molybdenum inclined at 20° with respect to magnetic field. The limiters were posi-tioned with their tips at the last closed flux surface (LCFS). The injection holes with a diameter of 1.7 mm were 15 mm behind the LCFS. After exposure in a series of reproducible ohmic discharges with 13CH4 injection, the plates were analysed ex-situ. No background 12C could be detected on tungsten, while on molybdenum ~ 1017 C/cm2 were found. On graphite the background deposition was about one order of magnitude higher. This is in agreement with the expected dependence of carbon deposition on the substrate mass. The deposition efficiency was found to be ~ 0.5 % for both molybdenum and tungsten limiters (see below, section on present modelling activities). The value for graphite is similar, being in agreement with previous observations made with roof-like carbon limiters.
Code development
The technical part of the parallelisation of the ERO code has been completed. First tests demon-strate a decrease of calculation time by a factor between 5 and 10 depending on the parameters of the specific simulation. Some code optimisation, especially concerning the transport of chemically erodes molecules, will be done.
Coupling of ERO with the Monte-Carlo code SDTrimSP has been finished. The SDTrimSP code simulates the transport of ions in matter including erosion and deposition and therefore delivers a dynamic change of the surface composition. In contrast to the former surface model of ERO it is now possible to get depth profiles of deposited species. Furthermore, effects resulting from the mix-ing of species with different atomic numbers can now be described better. The application of the coupled ERO version to 13CH4 injection experiments in TEXTOR using different substrate materials (graphite, tungsten and molybdenum) has been started.
Present modelling activities
Modelling for the linear plasma simulator PISCES (San Diego, USA) has been started. The ERO version for the PISCES version has been updated. A reasonable agreement between measured axial beryllium emission profiles and simulated ones can be obtained if collisions with the neutral gas are taken into account. From this it can be concluded that the transport of beryllium in the PISCES plasma is well described by the ERO simulations. Next step will be the simulation of deposition and erosion of mixed Be-C-W layers observed at the PISCES target plate.
Simulations of the hydrocarbon transport through nozzle-like geometries exposed into the edge plasma of TEXTOR show that the calculated D/XB values (‘dissociation per photon’ values to re-late observed light emission to particle fluxes) depend mainly on plasma parameters (decrease with increasing plasma density and increase with increasing plasma temperature) and puffing rate
Simulations of the hydrocarbon transport through nozzle-like geometries exposed into the edge plasma of TEXTOR show that the calculated D/XB values (‘dissociation per photon’ values to re-late observed light emission to particle fluxes) depend mainly on plasma parameters (decrease with increasing plasma density and increase with increasing plasma temperature) and puffing rate