Doc. SC-FZJ 78(07)/4.1.2
Nuclear Fusion Project
Association EURATOM / Forschungszentrum Jülich
A NNUAL P ROGRESS R EPORT 2006
including the contributions of the TEC Partners
ERM/KMS Brussels and FOM Nieuwegein
May 2007
edited by Ralph P. Schorn r.p.schorn@fz-juelich.de
A. The FZJ Nuclear Fusion Programme (executive summary) ... 5
B. Scientific and Technological Programme... 12
B.1. Plasma-Wall Interaction ... 12
B.2. Tokamak Physics ... 33
B.3. Diagnostics and Heating ... 61
B.4. Contributions to ITER ... 88
B.5. Contributions to Wendelstein 7-X... 103
B.6. Materials and Components under high Heat Loads ... 115
B.7. Theory and Modelling... 135
B.8. Oxydation Measurements ... 164
C. Specific Contributions of the Partners within the IEA Implementing Agreement... 180
C.1. Japan... 180
C.2. United States of America... 185
D. Summary on results of the main projects in the framework of "Projects for ... 190
enhancing the mutual co-operation between Associations" E. Structure of the FZJ Nuclear Fusion Programme and related Figures ... 197
F. List of scientific Publications, Talks and Posters ... 200
(executive summary)
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The FZJ Project Nuclear Fusion is part of the international nuclear fusion research, which pursues the long-term goal of realising on earth the method of power generation employed by the sun. Thus, a new and practically inexhaustible energy source with favourable safety and environmental charac- teristics shall be made available to mankind. The FZJ Institute for Energy Research / Plasma Physics (IEF-4) and Microstructure and Properties of Materials (IEF-2) perform the major part of research activities within the project. For more details please read chapter E: "Structure and Fig- ures of the FZJ Nuclear Fusion Project and related Figures".
The research programme of the project is oriented towards the strategy of the European research programme (Association EURATOM-FZJ and European Fusion Development Agreement EFDA), where the realisation of ITER, research in support for ITER and the construction of the stellarator Wendelstein 7-X in Greifswald – as the most promising alternative to the tokamak – play a central role. This report presents results having been achieved under the EURATOM Association Contract during the year 2006.
1 – Highlights 2006
ITER-like wall in the Joint European Torus (JET)
At the European fusion experiment JET (Abingdon, United Kingdom) a decisive hurdle has been overcome for the research area of plasma-wall interaction. Researchers from Jülich and Garching were responsible for the development of two concepts for lining wall components with tungsten.
Tungsten coatings, which have already been tested in the ASDEX Upgrade tokamak in Garching, are now also being used for the heavily loaded parts of the vessel wall – together with a new design for bulk tungsten plates developed at Research Centre Jülich.
The tungsten plates consist of a lamella structure with electrical insulation and predefined current paths on a web-like fixing structure. This configuration fulfils the special thermomechanical criteria for use in a tokamak. The design has already been successfully tested at Research Centre Jülich in the JUDITH electron beam facility and in the TEXTOR tokamak.
Industrial manufacturing has since begun. It is planned to incorporate the new wall coatings into JET in 2009. By extensively lining the main part of the burn chamber with beryllium and the diver- tor with graphite (CFC), we will be able to test the proposed combination of wall materials for ITER for the first time on a large scale.
Suppression of ELMs by Resonant Magnetic Perturbations (RMP)
The Dynamic Ergodic Divertor (DED), which was put into operation on TEXTOR in 2003, gener- ates externally applied resonant magnetic perturbation near fields in different modes which – so far
unique – can rotate with a frequency of up to 10 kHz. This provides a powerful experimental tool to study transport, energy and particle exhaust, as well as plasma stability issues and is of high rele- vance for plasma-wall interaction studies.
In 2006 a limiter H-mode has been generated in TEXTOR. By employing the DED this plasma re- gime is used to study the mitigation of peak power loads caused by ELMs. Moreover, recent results from JET performed under the leadership of FZJ scientists have also demonstrated the suppression of ELMs by RMPs. These results motivate further work to study the possibility for implementation of this technique also at ITER and will become an important element within joint experiments to- gether with DIII-D at San Diego, USA.
2 – Objectives and incorporation into the research area
Fusion research in Germany – organized within the Helmholtz Association of German Research Centres (HGF) and being in line with worldwide activities – follows the long-term objective to open the use of the nuclear fusion principle for the generation of energy on earth using the sun as a model – and thus securing humankind an abundant new primary energy source with favourable safety and environmental characteristics. Fusion energy could contribute to the generation of base-load power on a large scale starting from the middle of the century.
With the decision to build ITER in Cadarache (Southern France), the globally coordinated nuclear fusion research will attain the first burning fusion plasma with a power generation of 500 MW, a burn time of eight minutes and a tenfold power amplification factor. The ITER experiment together with the results of the accompanying research programme (materials development, fusion technol- ogy, advanced plasma physics) will be decisive for the construction of the first fusion power plant DEMO, which is intended for continuous operation and for energy supply into the grid. Thanks to its specific potential for continuous operation, the stellarator concept is considered an attractive al- ternative to the tokamak. The optimized Wendelstein 7-X stellarator in Greifswald, which is cur- rently under construction, will serve to explore the basic suitability of this concept.
The Helmholtz Association's fusion activities are based on the European fusion research pro- gramme. The following Helmholtz Centres are involved: Max Planck Institute of Plasma Physics (IPP, Garching and Greifswald), Research Centre Karlsruhe (FZK), and Research Centre Jülich (FZJ). In 2006, the former four research topics have been reorganized into six topics – ensuring that particularly the fields "plasma-wall interaction" and "theory" are now individually represented stressing their interdisciplinary character. The six topics are now as follows: a) stellarator research, b) tokamak physics – ITER and beyond, c) fusion technology for ITER, d) fusion technology after ITER, e) plasma-wall interaction, and f) plasma theory. Within HGF Research Centre Jülich is in charge of the topic "plasma-wall interaction" and is involved in all other topics except for the field of "fusion technology after ITER".
3 – Programme results a. Stellarator research
The fabrication of components for Wendelstein 7-X showed considerable progress in 2006. Jülich's contribution was concentrated on the superconducting bus-bar system and on plasma diagnostics.
The superconducting bus-bar system and the relevant joints were developed and manufactured in Jülich. This system consists of a geometrically complex mesh of conductors being exposed to strong forces. It provides the electrical connections to the stellarator's magnetic field coils. After the
construction of the production line and the qualification of the fabrication and assessment process, the first set of six bus conductors was manufactured and successfully tested. The construction and fabrication of supporting elements are currently being worked on. Series production was optimised concerning the joints. A welding procedure was developed and tested for repeated mountability.
The development and expansion of diagnostics is proceeding according to plan. The first experi- ments using the completed HEXOS VUV spectrometer have been conducted at TEXTOR. Agree- ments for the development of diagnostics have been reached with four more cooperation partners.
b. Tokamak physics – ITER and beyond
This research topic deals with experimental work performed on TEXTOR and on other tokamaks, such as JET and ASDEX Upgrade.
Active control of edge localized modes (ELMs)
With the involvement of Research Centre Jülich, experiments on ELM suppression (i.e. periodical plasma instabilities near the walls) were conducted at JET using external error field correction coils.
They succeeded in reducing the ELM amplitude to a quarter. Furthermore, the energy loss that oc- curs with the appearance of an ELM could be reduced distinctly, which is extremely important in terms of extending the lifetime of the first wall.
At TEXTOR the basic characteristics of particle and energy transport have been investigated in a stochastized edge plasma that was generated using interference fields produced with the aid of the Dynamic Ergodic Divertor (DED). Thus, a similar reduction in ELM activity could be observed also at TEXTOR.
Detailed studies on plasma transport revealed the relevance of what is known as the "laminar zone"
showing short connection lengths to the divertor as an important plasma sink. Operating with a DED coil configuration of m/n = 6/2 was particularly suitable for these studies because it allows a maximum stochastization without the stimulation of instabilities in the plasma centre, for instance tearing modes. We also succeeded in showing that stochastization increases the extraction of fast particles ("runaway electrons"), which could be potentially important for the protection of the burn chamber walls.
Avoiding and minimizing disruptions
Disruptions caused by instabilities in ITER could lead to stresses on the vessel structures as a result of highly energetic electrons, uncontrolled thermal fluxes and electrical currents inside the struc- tures. Studies on minimizing these effects with massive gas injections are being conducted at TEX- TOR and ASDEX Upgrade. For this purpose different gasses, such as helium, argon and mixtures with deuterium were injected into the plasma using an ultra-fast valve developed at FZJ. It was shown that injecting argon leads to the formation of runaway electrons. These can be avoided using a mixture of 10 % argon and 90 % deuterium. With respect to the mixture, however, it is difficult to control the density in the following plasma discharge. When helium was injected the formation of runaway electrons was not observed and the consecutive discharge was unaffected. The same type of valve was integrated into JET; experiments with the involvement of Research Centre Jülich have been prepared in order to allow a "disruption mitigation system" (DMS) be scaled for use at ITER.
c. Fusion technology for ITER
Materials for components in contact with the ITER plasma
Work at Research Centre Jülich is concentrated on the characterisation of actively cooled modules for the first wall and the divertor in ITER as well as on passively cooled wall elements for an ITER- type wall in JET. Apart from the determination of the mechanical and thermophysical material characteristics, fatigue behaviour under cyclic thermal loading is of special interest. In 2006, mono- lithic CFCs (NB31) and sintered tungsten were investigated together with layered systems (tungsten on CFC and beryllium on metallic substrates). In order to extend the loading parameters and test capacities, the new electron beam test facility JUDITH 2 was put into operation at FZJ.
Development of diagnostics and port plugs for ITER
New measurement technologies for ITER have been developed at TEXTOR in the areas of the laser desorption of deuterium and tritium, desorption quantification using a quartz micro balance and dispersion interferometry. Under the direction of Research Centre Jülich and its Netherlands TEC partner FOM, a consortium was set up for the development of the "ITER CXRS diagnostic port plug". The objective is to use charge exchange recombination spectroscopy (CXRS) to examine the ITER plasma. The physical design of the system is almost fully completed. Accompanying research and development projects were conducted in the following areas: testing of materials for the first mirror at TEXTOR; preparation of a benchmark experiment at TEXTOR on charge exchange re- combination spectroscopy with ITER-relevant geometry and mechanical strength studies ("Port Plug Engineering"). Work has been started on the development of a project plan for the design and construction of the ITER CXRS port plugs.
d. Plasma-wall interaction
Controlling the interaction between the fusion plasma and the walls of the burn chamber is becom- ing increasingly important for ITER's operating range and for further developments leading to a steady-state burning fusion plasma. The aim here is to attain conditions with very little erosion of the wall material and at the same time with a very low fuel retention. In addition, these conditions must be compatible with the requirements for energy confinement in ITER. Reaching this goal is strongly influenced by the choice of wall materials: For ITER e.g. a combination of carbon resp.
graphite, beryllium and tungsten is foreseen at the moment. For experimental research performed until now, the JET, TEXTOR and ASDEX Upgrade tokamaks have been the most prominent facili- ties available – together with various devices within HGF and those belonging to international part- ners.
Experiments on erosion, deposition, carbon migration and fuel retention
In JET, we could show that erosion and deposition behaviour are strongly dependent on the mag- netic field geometry: Deposition primarily occurs in direct line of sight of the erosion source; the inner divertor is dominated by deposition, while the outer divertor regions can be dominated by either erosion or deposition. The appearance of edge localized modes (ELMs) significantly in- creases the amount of material deposited.
In TEXTOR the retention of deuterium in different carbon materials was found to be very similar for CFC and fine-grained graphite (difference of about 30 %) in case of high particle flux densities.
No saturation effect was observed. Experiments with an ITER-like castellation of plasma-limiting materials at TEXTOR, ASDEX Upgrade and DIII-D showed that fuel retention in cracks can be reduced by both a higher temperature and an optimised tile form.
One of the methods for removing deposited hydrocarbon layers is oxidation in an O2/He glow dis- charge. In ASDEX Upgrade and TEXTOR, it was shown that this process only removes layers of pure hydrocarbons efficiently from the vessel walls, while layers that have been mixed with boron (for instance from boronisations) simply oxidise and are only eroded to a very small degree.
High-Z wall materials and mixed material systems
After the last experimental period, also the remaining surface areas and divertor plates have been coated with tungsten in ASDEX Upgrade. This made a complete wall covered by tungsten available for research in a tokamak.
In TEXTOR, experiments with high-Z materials have been continued employing extreme thermal fluxes. The here observed increase in tungsten erosion can be explained by the sputtering of the material as a result of bombardment with carbon particles and also as a result of normal sublima- tion. By imposing maximum thermal loading the behaviour of the tungsten melt was analysed. It was shown that the interaction between the thermally emitted flow of electrons and the magnetic field is the driving force for the observed movement of the melt. The brittle fracture behaviour of tungsten in the case of transient events was investigated on the JUDITH electron beam facility at Research Centre Jülich and on plasma jet facilities in Russia. JUDITH was used for further basic research on the determination and comprehension of limit values for crack formation and phase transition processes, as well as for the realization of thermally induced material changes as a func- tion of the number of cycles of transient loads.
For the lining of the new JET divertor ("ITER-like walls"), research was continued on the qualifica- tion of bulk tungsten tiles and tungsten coatings. At Research Centre Jülich, a lamella concept com- patible with the electromagnetic forces in JET was developed employing bulk tungsten tiles that was successfully tested in JUDITH and TEXTOR.
Modelling
The modelling of erosion, deposition and impurity transport using the ERO code was improved: By combining ERO with a surface model based on the TriDyn code, we are now able to describe ero- sion in mixed material systems. The decomposition of different hydrocarbons was determined by means of experiments at TEXTOR and results have been introduced into the codes. The influence of beryllium – a material that is also intended to be used in ITER – in mixed material systems will be investigated experimentally using the PISCES linear plasma facility (USA) and will also be modelled using the ERO code.
Diagnostics
A new method for the in-situ determination of the hydrogen content of carbon layers using laser desorption and spectroscopy has been developed at TEXTOR. At JET, spectroscopy on hydrogen isotopomers was used to determine low concentrations of hydrogen or tritium against a high back- ground of deuterium. Comparative experiments on the erosion and deposition behaviour of optical mirrors for ITER diagnostics were conducted in TEXTOR, ASDEX Upgrade, and other fusion- relevant experiments. They revealed the importance of the operating temperature of the mirror.
Work continued on the development of purification processes, e.g. using glow discharge, ECR, and ICR plasmas. The spectroscopic determination of the tungsten release from wall components was also expanded.
e. Theory
Plasma edge physics
Among other topics, Research Centre Jülich is engaged in the kinetic modelling of neutral gas, ra- diation transfer, and kinetic effects on the transport behaviour of charged particles in the edge re- gion. In the EIRENE code, with which the Boltzmann equation and the Fokker Planck equation can be solved in the plasma edge region and in the divertor, the atomic data basis was supplemented with hydrocarbons; nonlinear effects have also been added, e.g. friction and viscosity inside the neutral gas flow, the Zeeman and Stark splitting of the spectral lines, and drifts of charged particles in electric and curved magnetic fields. The extended code was integrated into the ITER team's simulation package. It is also employed now for design studies addressing ITER and DEMO in co- operation with ITER colleagues. In particular, research is currently being conducted on an option for simplifying the divertor design, especially concerning the plasma and neutral gas fields that have been altered due to effects that have to be taken into account additionally.
Numerous 2-D simulation calculations conducted at the Max Planck Institute of Plasma Physics in Garching using the B2-EIRENE code showed a systematic tendency to overestimate the density in the divertor and to underestimate the temperature for discharges in ASDEX Upgrade. Neutral parti- cle effects as well as kinetic effects have at first been suggested as possible reasons for this. Since then however, everything seems to point towards the latter as the main cause. The parallel heat transport in the plasma's scrape-off layer is apparently not adequately described with a fluid model.
The integrated simulation of 3-D plasma edge regions employing the EMC3-EIRENE code has been validated by conducting a detailed comparison with TEXTOR-DED experiments. Thus, the predictive quality was significantly improved. Since the complex topology of the magnetic field could be eliminated as an uncertainty factor by directly coupling the transport code with precise magnetic field calculations using "field line tracing" (Gourdon code), we were able to identify criti- cal model assumptions being relevant for a quantitative analysis. As was the case for the above- mentioned two-dimensional (i.e. axially symmetric) simulations on ASDEX Upgrade, the focus here too lies on kinetic effects in the plasma simulation. In addition, the sputtering model (yield, energy distribution) and boundary conditions at the interface to the plasma centre were identified as sensitive parameters for impurities. Corresponding development on new codes has been started with which relevant model components can now be calculated "ab initio" within the integrated transport model.
Semi-empirical transport modelling
A new model for the representation of particle and energy losses from the plasma was worked out for "type I edge localized modes" (ELMs). Here, transport in the case of ELMs is increased due to fluxes along the field lines, which are disturbed by MHD instabilities. Thus, the experimentally determined dependency of losses on the collisionality of the plasma can by explained well. ELM suppression using external coils in the DIII-D tokamak (San Diego, USA) could also be described within the framework of these assumptions. In addition, also kinetic effects play an important role in transport behaviour in terms of the power threshold.
Using the RITM code, the behaviour in the case of the L-H transition – being different in divertor and in limiter tokamaks – could be explained by the contribution from convective thermal losses which are much larger in limiter machines. This model was implemented into the JETTO code at JET. In this way, the experimental dependence of the power threshold on the magnetic field strength and on the density of the plasma could be reproduced well for this transition.
Turbulence
The impact of ergodized magnetic fields on plasma confinement and in particular on edge localized modes (ELMs) is being investigated in Jülich with the ATTEMPT turbulence code. It was shown that in the simulation radial plasma transport conformed to observations made in TEXTOR's edge region. After the ATTEMPT code was extended to include a scrape-off layer model, what is known as "blob transport" became visible in the simulation. This means that the plasma intermittently en- teres the scrape-off layer in blobs. By switching on an interference field, this blob transport could be heavily suppressed, which qualitatively corresponds with turbulence measurements in the TEXTOR edge region.
f. Cross-programme topics
The range of problems associated with the increased demand for nuclear fusion experts, particularly in the field of engineering – caused by ITER itself and also by the new support organisations which have to be built up ("European Joint Undertaking for ITER", "Broader Approach" with Japan) – has meanwhile been taken note of by the European Commission. An invitation to tender for nine Euro- pean training networks subsequently followed, in which three German Helmholtz Centres who are involved in fusion research have been extremely successful: They are now involved in eight net- works – four of which they are coordinating – and will take on and train a total of 18 young scientist and engineers, four of whom will work in Jülich.
To a large extent, the nuclear fusion programme is kept alive by intensive cooperations with a wide range of partners. This is the basis upon which the new linear plasma facility MAGNUM-PSI, cur- rently being built at the FOM Institute (our TEC partner) in Rijnhuizen, The Netherlands, will even- tually be run. This steady-state high flux facility will make a powerful new tool available for plasma-wall interaction research, particularly for issues that could so far not be detected in pulsed tokamak plasma facilities. Research Centre Jülich is involved in building the facility through the construction of the target station and it will also be involved in utilising MAGNUM-PSI later as a partner.
As well as research institutions being associated with EURATOM, partners from Jülich's point of view also include universities (Bochum, Düsseldorf, Duisburg-Essen, and a number of universities in Belgium and The Netherlands) as well as institutions in the "new" EU member states, whereby scientific relationships between the "old" and "new" member states are supported in the EURA- TOM fusion programme with special emphasis. Last but not least an intensive exchange exists with Russia.
Cooperations with the EURATOM Fusion Associations will be put on a new basis for the erection of ITER. At the moment, European consortia of Associations are being formed to take over the re- sponsibility of implementing individual systems and components from the European package of tasks for ITER. Complex systems and constructions are involved here, requiring the specific know- how of the individual fusion laboratories.
B.1. Plasma-Wall Interaction
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Introduction
The decision to build ITER in Europe resp. France has recently been endorsed. Providing solutions to remaining physical difficulties and open scientific questions, especially to those related to plasma-wall interactions, is thus even a more urgent need than before – in order to be able to make the best engineering choices. The TEC Main Topic Group on Plasma-Wall Interaction (PWI) re- mains organised in a programme-centric fashion and, with the motivation of developing a viable solution to the questions of the ITER first wall and divertor materials, the group is fully aligned with the goals of the European Task Force on PWI. For details see
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ITER-relevant subjects are thus treated with priority. JET will, for instance, study relevant mate- rial combinations in the frame of the ITER-like wall project. In parallel, the basic understanding of plasma-surface interaction and related plasma processes near plasma-facing components is still to be improved.
The two categories of open questions for PWI on ITER are clearly: (i) measurement and prediction of fuel retention on the one side, and (ii) the lifetime of the plasma-facing components (PFC) on the other side, especially of the target and other highly exposed elements, a question related to the ero- sion processes. The qualification of high-Z materials in a form suitable for plasma-facing compo- nents accordingly is an important subject as well.
Thus, research in the Main Topic Group on Plasma-Wall Interaction in first instance addresses the following 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 deposition mitigation techniques and of removal techniques for hydrogen isotopes belongs to this field, hereafter referred to as section 1.
• High-Z plasma-facing components and mixed materials. These are possible solutions for the baffles or for the case of an extensive tungsten covering in ITER, possibly even including the divertor plates. This topic is addressed in section 2.
The Plasma-Wall Interaction Group performs work on a large number of fusion facilities in order to match best the needs of any specific physical investigation: Besides TEXTOR, also JET, AUG (AS- DEX-Upgrade), DIII-D, PISCES, Pilot-PSI, and others are employed. The full research programme is organised within the Trilateral Euregio Cluster (TEC), which encompasses the Belgian (ERM/KMS) and Netherlands (FOM) partners in addition to the German FZJ institute at Jülich. The
IAE partners (Japan, USA, Canada) are closely linked to this research programme as well, as can be seen in the following. TEXTOR serves as the central fusion facility for the TEC partners, without prejudice to resorting to any other device if better suited.
1. Erosion and deposition, carbon migration and tritium retention 1.A. Chemical erosion, hydrocarbon catabolism and spectroscopy
Carbon-based materials have been deployed successfully as plasma-facing materials in present-day fusion devices. Best plasma performance was achieved with carbon-fibre components (CFC) used as divertor target plates at the locations of highest heat flux. However, chemical erosion of carbon and the appearance of co-deposition of fuel are critical issues in the usage of carbon-based plasma- facing components and leads, in the case of tritium, to the limited application of CFC in the ITER divertor which is presently foreseen. The main question is if the remaining CFC is acceptable in the tritium phase of ITER operation or not. This topic is closely related to the determination of chemi- cal erosion yields under ITER-relevant plasma conditions because a substantial part of the long term tritium inventory is found in the carbon-based re-deposited layers. Quantification of CFC erosion yields and their estimation with erosion and deposition simulation codes, e.g. ERO (see below, sec- tion 1.H.), is the topic of the ITPA Divertor and Sol Group (D-SOL2) and of the SEWG (Special Expert Working Group) on Chemical Erosion in the EU Task Force on Plasma-Wall Interaction.
Experiments under the scientific leadership of FZJ have been performed at JET, TEXTOR and AS- DEX Upgrade. These experiments focus on
a) the data base (HYDKIN) and code validation (ERO-code), and
b) the measurement of erosion yields in situ, in dependence on energy and flux of incident particles as well as on surface conditions (validation of the so-called "Roth formula").
The quantification of chemical erosion yields is mainly based on the detection of hydrocarbon break-up products such as C2 or CH by spectroscopy. This is done through injection of a well- known amount of various hydrocarbon gases. The ratio of particles injected into the plasma – which are later on dissociated or ionised – to the amount of light of the break-up products at the end of the chain provides effective efficiency factors. The C2 Swan band and the CD Gerö band are commonly observed for this purpose.
→ At TEXTOR, a series of experiments has been carried out to determine these inverse photon efficiencies, in particular of the pairs CH/Gerö band and C2/Swan-band and, thus, to verify the HYDKIN database and the plasma-related part of the ERO code. These experiments were per- formed with gas injection modules made of metal and minimised contact surface area to exclude any surface effects. Photon efficiencies for typical ohmic TEXTOR plasmas with electron tempera- tures of about 50 eV at the LCFS as well as HYDKIN calculations are given in Table 1. The data are in a fair agreement considering the assumption of a constant plasma background in HYDKIN.
Figure 1 provides an overview of measured and calculated photon efficiencies for methane and the full emission of the Gerö band. Deviations might be attributed either to the simplifying assumption of a constant plasma background in HYDKIN, or to surface effects which cannot be excluded for all ex- periments in TEXTOR.
Fig. 1: Measured effective D/XB values for methane deduced in different experiments and compared with HYDKIN calculations. Spectroscopic data
is normalised to the full emission of the Gerö band.
Table 1: Measured (TEXTOR) and calculated (HYDKIN) effective D/XB values for hydrocarbons up to the propane group.
→ Experiments at ASDEX-Upgrade focussed on the determination of the chemical erosion yield under detached divertor conditions which are comparable to the regime foreseen for ITER. A dis- charge scenario containing strike-point sweeps was developed in order to detach the outer divertor in L-mode with the aid of high deuterium fuelling. A strong reduction of the intrinsic CH Gerö band emission was observed when the divertor detaches. Injection of C2H4 or CH4 leads to a significant extrinsic Gerö band photon flux. The corresponding inverse photon efficiencies are larger than the theoretically predicted ones (HYDKIN) and slightly larger than the ones previously deduced in high density discharges with attached divertor leg (see figure 2).
Fig. 2: Strong decrease of the CD Gerö band when the divertor is switched from the attached to the detached plasma regime. Injection of methane lead to a significant increase of the CH radical produced during the
hydrocarbon catabolism. The difference in the overall intensity of the spectrum is opposite to the theoretical prediction from the HYDKIN code.
The behaviour of the C2 Swan band was comparable, but the intrinsic light emission was below the detection limit of the applied spectroscopic system. The deduced intrinsic hydrocarbon flux was reduced by one order of magnitude in comparison with the attached reference case. However, this strong reduction is partially compensated by the reduction of the impinging ion flux when the diver- tor detaches. The erosion yield itself is only reduced if the particle flux of neutral atoms to the target is taken into account. These experimental results confirm the Roth formula for detached plasma conditions if both impinging fluxes, the atomic and the ion flux, are considered.
The consequences for ITER can be described in the following way:
i. Detachment of the outer divertor actually leads to a reduction of the hydrocarbon flux.
ii. The overall erosion of graphite at the outer target plate is reduced. Note that the outer diver- tor was recently identified to be the remaining carbon source in ASDEX-Upgrade with tung- sten main chamber.
→ Several experiments in JET were performed to investigate the appearance and the disintegration of a-CH layers in dependence on the ELM energy, the strike-point configuration and the history of previously performed shots – labelled "history effect". QMB (quartz micro-balance) measurements and hydrocarbon spectroscopy were applied as diagnostic methods to monitor the gross and net- erosion – and the inner strike point was utilised as a tool to detect the location of the layer. The lo- cation of loosely bound layers in the inner divertor was found in most cases close to the corner re- gion (Fig. 3). Discharges in H-mode with type I ELMs and fast strike-point sweeps over a few cen- timetres on the horizontal base plate were applied to disintegrate the layer. Figure 3 shows the deposition of hydrocarbons onto the QMB in red bars when disintegrating the soft a-CH layer. The
blue (type III ELMs) and green bars (type I ELMs) represent the deposition onto the QMB as a function of the ELM type for the same strike-point configuration.
Fig. 3: Deposition of hydrocarbon onto the QMB as a function of the ELM type (JET inner divertor).
1.B. Quartz Micro-Balances in JET
A quartz micro-balance (QMB) set-up can monitor deposition processes, as the resonance fre- quency of the quartz crystal depends on the amount of material deposited on its surface. Six QMB systems were installed during the last JET shutdown: QMB1, QMB2 (heated) and QMB3 (cooled) in the inner divertor, QMB4 in the private flux region, QMB5 (heated) in the outer divertor and QMB6 in the outer apron region to monitor the beryllium evaporation. Fig. 4 shows the data col- lected with QMB1 and QMB5 during the restart and JET experimental campaigns C15, C16 and C17 with some information on the experimental geometry and the QMB mass calibration.
Fig. 4: Overview of the database of the QMB1 (inner divertor) and QMB5 (outer divertor) systems.
The highest deposition rates were obtained in one of the H-mode cleaning discharges (JET pulse
#68133) with large ELMs and with a sweep of the strike points described in the previous section.
The analysis of the QMB database is ongoing. The main results obtained so far are:
- The deposition and erosion behaviour in the divertor is strongly influenced by the magnetic field configuration, e.g. by the positions of the inner and outer strike points.
- The inner divertor is dominated by deposition, whereas the deposition/erosion behaviour of the outer divertor varies with the discharge parameters.
- The amount of the deposited material in the inner divertor depends strongly on the presence of ELMs and on their size.
1.C. Carbon erosion measured in Pilot PSI
Carbon samples have for the first time been exposed to high fluxes in the Netherlands Pilot PSI linear plasma generator, the forerunner of MAGNUM. The device delivers hydrogen plasmas (0.5 – 6 eV) with fluxes up to about 5×1024 m-2s-1 in magnetic fields up to 6 T. High erosion rates were attained in this experiment, which is summarised in Fig. 5. This gives confidence that the expected and desired parameters for MAGNUM can be reached.
Fig. 5: The Pilot PSI device and an example of a carbon erosion experiment.
1.D. Retention in Carbon-Fibre Composites (CFC) compared to graphite
Three different carbon based materials (CFC NB 31, CFC DMS780 and fine-grain graphite EK98) have simultaneously been exposed in the SOL of TEXTOR to fluences of up to ~ 2×1025 D/m2 in order to measure the retention of deuterium (Fig. 6). The thermo-desorption analysis of the samples showed a similar deuterium retention for both CFC materials and a retention being lower by about 20 to 30% for EK98. The amount of retention of (2–3)×1021 D/m2 for the highest fluences is in a good agreement with the available database. No sign of a saturation was observed in the retention.
An important result from the estimations for TEXTOR is that the amount of D retained in graphite (and similarly in CFC) is by a factor of about 3 lower than that of deuterium co-deposited in a-C:D layers in the same period of time.
Fig. 6: Test limiters with the different carbon/CFC stripes after exposure.
1.E. 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 to 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 have been determined in L- and H-mode shots. The way to determine the D/T flux ratio via the measurement of atomic lines (preferentially Balmer α) encloses several difficulties. Because of their dependence on the magnetic field and on the width (resp. energy distribution) of the individual components the spectra have to be de-convoluted, which may be a tedious task with several unknowns. An example of such a spectrum is shown in Fig. 7.
Fig. 7: Balmer α lines emitted from D/T atoms at the location of the T2-puff in JET.
This favours the use of molecular spectra. Therefore, it was tried to measure the D, H and T fluxes in D discharges via the observation of the isotopomers HD and TD. This method provides a very sensitive way to detect a low minority concentration (H or T) versus a high deuterium background
level. 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 (see report 2005). Fig. 8 displays how well the spectra of deuterium and tritium molecules can be separated as the respective lines are emitted at quite different wavelengths and do not show detectable Zeeman broadening. From the line intensities a tritium to deuterium ratio of one can be deduced at the location of the T2-puff. From the recording of the TD spectral line emission a recycling T2 flux could be derived being equal to about 0.02 % of the D2 flux.
Fig. 8: Molecular line emission from deuterium and tritium molecules at the location of the T2 puff.
1.F. Investigations of ITER-like castellation structures in the TEXTOR tokamak
Investigations of ITER-like castellation structures are underway at TEXTOR to study carbon migra- tion and fuel deposition in gaps. In a recent experiment, a new tungsten castellated limiter has been exposed. Castellation cells of rectangular and of roof-like geometries as shown in Figure 9 were exposed to the same plasma conditions in the SOL of TEXTOR, allowing a direct comparison of the carbon transport and fuel inventory in the gaps. After exposure, the cells were dismantled and the deposits on the gap sides were studied using several surface analysis techniques: Secondary Ion Mass Spectrometry (SIMS), Nuclear Reaction Analysis (NRA), Electron Probe Micro-Analysis (EPMA) and Dektak surface profiling.
Fig. 9: A Tungsten castellated limiter after exposure to the TEXTOR plasma.
The gaps of shaped cells contained a factor of 2 to 3 less carbon and deuterium in comparison to those of rectangular cells (see Fig. 10). This fact is evidencing the advantage of using adequately shaped castellation blocks.
Fig. 10: Deuterium (fuel) atomic concentration along the depth of the gaps for a) the plasma-open side, and b) for the plasma-shadowed side of poloidal gaps for different cell geometries.
Future studies will focus on the quantification of the fuel amount in the gaps, on the quantification of metal intermixing in the deposits, on a further optimisation of the cell shape, on the analysis of carbon transport and fuel inventory of toroidal gaps, on the comparison between toroidal and pol- oidal gaps and on the modelling of carbon deposition and fuel co-deposition in the gaps. The inves- tigations are carried out in the framework of multi-machine ITPA (Task DSOL-13) of the ITPA TG on Divertor Physics and SOL.
1.G. Erosion/Deposition: Investigation of diagnostic mirrors for ITER
Studies of ITER candidate mirror materials and technological concepts have been continued in the framework of the High Priority Task of the ITPA TG on Diagnostics. Currently, the research of diagnostic mirrors is a part of the multi-machine ITPA. Maintaining the optical properties (reflectiv- ity) of mirrors in spite of high erosion or impurity deposition during ITER operation is indeed a major challenge.
A series of mirror surface finishing tests was performed at TEXTOR in collaboration with the Uni- versity of Basel (Switzerland), the Kurchatov Institute (Russia) and ENEA Frascati (Italy). These studies were aiming at assessing alternative mirror solutions to be used for ITER in the erosion en- vironment. One of the promising alternatives is the use of the mirrors with highly reflecting coat- ings. Molybdenum mirrors coated with nano-crystalline films and rhodium coated mirrors from various manufacturers were exposed under the same erosion dominated conditions – with single crystal molybdenum mirrors as a reference. On the basis of these investigations the conclusion has been drawn that the technology of coating should be optimised in order to increase the sputter resis- tance of coated layers.
The cleaning of the mirror exposed under deposition dominated conditions in the SOL of TEXTOR has been performed at the IPP Kharkov using ECRH discharges in deuterium. For the majority of studied mirror locations, it was possible to restore the initial reflectivity of the mirror after having been degraded during exposure to the TEXTOR plasma.
A new dedicated test of diagnostic mirrors in a tokamak divertor was performed at the National Fusion Facility DIII-D (General Atomics, U.S.A.) under the lead of specialists from FZJ. The mir- rors were exposed in the DiMES transport system below the newly installed divertor floor during 8
plasma discharges, being similar to those used during the first two exposures in 2005. The total plasma duration was 36 seconds.
Fig. 11: Left: The limiter with single crystal and coated mirrors during exposures in the SOL of TEXTOR.
Right: The evolution of the total reflectivity of molybdenum mirrors before and after exposure in the DIII-D divertor: (1) before the first exposure at room temperature, [to be compared with] (2) after the first exposure. (3) before the second exposure at elevated, but drifting temperature, [to be compared
with] (4) after the second exposure. (5) before the third exposure at elevated and almost stable temperature, [to be compared with] (6) after the third exposure.
During the exposure, the temperature of the mirrors was kept at (150 ± 7) oC. After exposure no visible deposition was noticed, and the subsequent elemental surface analysis has confirmed an ab- sence of carbon on the mirrors – only negligible traces were found. Moreover, during optical meas- urements no decrease of optical reflectivity was detected, meaning that both goals of this investiga- tion – mitigation of carbon deposition and conservation of the reflectivity – were successfully at- tained during the last experiment. The respective dependences of the total reflectivity measured before and after each of these three exposures are shown in the right diagram of Figure 11. This work has been performed within the framework of the bilateral U.S.-EURATOM exchange pro- gramme.
A predictive modelling of mirror performance in ITER has been initiated. The B2-Eirene Monte Carlo plasma neutral code is being applied to reconstruct the edge parameters of ITER. Modelling was performed for the first mirror of the ITER core Charge eXchange Recombination Spectroscopy (CXRS) diagnostic. The first results of modelling coupled with an assessment of the maximum total erosion rate imply that erosion should not be a concern if the first mirror of the CXRS system is made of a single molybdenum crystal. Subsequently, the ERO Monte Carlo impurity transport code will be applied to evaluate the erosion and deposition patterns for the two EU-credited diagnostics:
CXRS and the LIght Detection And Ranging (LIDAR) system.
Investigations are partly supported by the EFDA contracts TW4-TPDS-DIADEV, TW5-TPDS- DIADEV and TW5-TPDS-DIARFB.
1.H. ERO Modelling of erosion and deposition
The ERO code, coupled with SDTrimSP – a code to simulate transport of ions in solid state mat- ter – has been used to model the local deposition of 13C resulting from 13CH4 injection through spherical graphite and tungsten test limiters. The observed 13C deposition on the tungsten test lim- iter is much more localised compared to the graphite limiter. In addition, the 13C deposition is about a factor of 10 smaller on tungsten. ERO simulations with a simple material mixing model do not
show any dependence of the 13C deposition pattern and efficiency on the substrate limiter material in steady state. As soon as a carbon layer is built up the underlying substrate material does not af- fect the further deposition. Using the SDTrimSP model instead reduces the 13C deposition effi- ciency on the tungsten limiter and this also leads to a more localised deposition pattern. It is seen that carbon particles from the injection, with energies in the eV range, are re-deposited near the sur- face and therefore can be sputtered more effectively. However, the modelled difference between tungsten and graphite is not yet as pronounced as observed in the experiment. The remaining discrep- ancy possibly is caused by surface roughness effects – in the experiment the graphite test limiters had a larger roughness than the tungsten ones.
The modelling of experiments with the linear plasma simulator PISCES-B has been improved by taking into account elastic collisions between neutral beryllium particles and deuterium molecules.
Moreover, the description of ionisation and excitation processes has been improved by considering the dependence of the corresponding rate coefficients on the electron density, in addition to the electron temperature dependence. The database for ionisation and excitation rates has been updated by using data from ADAS. With these improvements the observed axial profiles of beryllium light emission (Be I and Be II), resulting from seeded beryllium, can be reproduced much better with ERO. Without neutral collisions the modelled profiles are peaked around the location of seeding.
They become broader, as expected, due to the collisions with neutrals.
First simulations of the observed mitigation of chemical erosion in Be seeded plasmas indicate the importance of carbide formation for the mitigation process. The observed quadratic dependence of the characteristic time of mitigation on the Be concentration in the plasma cannot be explained by a simple dilution of the carbon concentration in the surface due to beryllium co-deposition. Non- linear effects arise from Be2C formation at the surface. Carbon which is bound in a carbide mole- cule seems to be protected against chemical erosion. This finally results in an increased mitigation of chemical erosion. Further calculations are planned using the coupled version of ERO and SDTrimSP in order to describe the carbide formation and the mixing of C and Be at the surface in a better way.
Modelling of the net erosion of a tungsten stripe at the outer divertor of JET has been performed.
Plasma parameters for the simulations are estimated from "campaign averaged" data. The modelling shows that the tungsten erosion is dominated by carbon impurity sputtering whereas the erosion due to deuterium is negligible. About 50% of the sputtered tungsten atoms are re-deposited on the tung- sten stripe. The majority of sputtered tungsten atoms are lost into the PFR originating from a very narrow region around the strike point. With the deuterium fluence distribution taken from calibrated Dα photon signals of the corresponding experimental campaigns and using the ERO results on re- deposition and particle recycling, the net erosion profile is modelled quantitatively. The lower limit of the maximum net erosion from the measurement is about 3 µm being in fair agreement with the calculated value of about 7 µm in the maximum.
The erosion of the beryllium start-up limiter in ITER has been estimated. In a first attempt a specific time point of the ramp-up phase (Fig. 12) has been used to calculate the erosion and re-deposition of beryllium. It is seen that only about 20% of eroded beryllium is re-deposited on the limiter. There- fore, the gross erosion is not significantly reduced by re-deposition. The maximum net erosion can be estimated to be about 1 nm/s. Additional simulations for other time points of the ramp-up and ramp-down phase are planned.
Fig. 12: Erosion of a beryllium limiter during the ramp-up phase of ITER.
New simulations have been carried out to estimate the tritium retention and target lifetime in ITER.
A beryllium influx into the divertor has been taken into account. Due to the fact that the erosion of the beryllium main wall and the transport of sputtered beryllium into the divertor are not very well known, two values of the beryllium concentration in the divertor plasma (0.1% and 1%) have been analysed. For comparison also the effect of a carbon instead of a beryllium influx has been ana- lysed. The carbon flux is set to 1% (carbon suffers from chemical erosion wherefore the carbon influx into the divertor should be larger than a beryllium influx). Chemical erosion of graphite is calculated according to the "Roth formula" taking into account the dependence of the deuterium flux, incoming deuterium energy and surface temperature. The erosion of re-deposited carbon spe- cies is assumed to be enhanced by a factor of ten compared to substrate carbon and the effective sticking of hydrocarbons is assumed to be negligible. These assumptions are based on modelling of injection experiments in TEXTOR. It is seen that the target lifetime (in the worst case about 7000 ITER discharges) is much less critical than the long-term tritium retention. The rate of the latter varies between 6 and 32 mg tritium per second with the lowest value for the case of 0.1% Be and the largest one for a carbon influx. Future modelling will include the improved description of mate- rial mixing in the surface with SDTrimSP.
1.J. ICRF Wall Conditioning in TEXTOR
Oxidation treatment of the vessel walls and tokamak post-oxidation recovery
Tritium removal from amorphous tritiated carbon layers is one of the major problems for a fusion reactor since the amount of in-vessel retained tritium is safety limited. The oxidation of the carbon layers by a plasma-assisted technique is considered as one of the promising tools. The oxidation treatment of the TEXTOR vessel was performed with pulsed ICRF discharges (τRF ≈ 3−8 s) in a He and O2 gas mixture and analysed in terms of ratios of the RF pulse length to the O2 puff duration.
The new scenario of ICRF Discharge Conditioning (ICRF-DC) was tested with a higher O2 injec-
tion rate after termination of the RF power. The increased output of the O-containing products (O2
by about 4 times and CO by roughly 2.5 times) was achieved in a reliable regime of the ICRF an- tenna operation (Fig. 13). The independently developed post-oxidation wall cleaning procedure as a successive set of the ICRF discharges in deuterium (to remove the O-containing products) and in helium (to de-saturate the walls with deuterium) resulted in a successful recovery of TEXTOR to normal operation.
Fig. 13: ICRF O-treatment at longer (a) and shorter (b) RF pulse length compared with the O2–puff duration (PRF ≈ 50 kW, f = 29.0 MHz, BT = 2.3 T, He+O2 gas mixture).
2. High-Z materials for plasma-facing components
2.A. Atomic and molecular data: the case of atomic tungsten
To convert measured spectroscopic line intensities to fluxes and densities, the conversion factor S/XB (ionisations per photon) has to be known. Various theoretical codes exist to determine these values – like GKU, R-matrix and databases (ADAS) etc. TEXTOR offers an excellent possibility to compare those theoretical data with experimental results.
For tungsten the values for energies and radiative transitions have been compiled from a recent ver- sion of the NIST database. Unfortunately, the level identifications are given for the three lowest configurations (not all terms) only. The ionisation rate coefficients were recalculated by ATOM for the lowest configurations 5d46s2 and 5d56s. Calculations of the excitation cross sections meet prob- lems due to a complicated coupling scheme and configuration mixing. For many levels the identifi- cation is unknown and, therefore, the semi-empirical Regemorter formula was used with the coronal approximation for excitation only from the group of "ground" levels. Fig. 14 displays a (reduced) level diagram of W I, with the lines observed in the edge plasma of TEXTOR from tungsten plates and limiters inserted through a lock.
Line ratios for transitions underlined in red are subject to electron temperature variations, those in green depend on the population of the ground state which may be affected by the temperature of the emitting tungsten surface. Moreover, S/XB values (the proportionality factor to convert photon into particle fluxes) for several W I lines in the visible and UV could be calculated in order to have a larger flexibility in the choice of the wavelength, thus avoiding high transmission losses in light fibres which may occur in the case of the line at 400.8 nm – the most used one in plasma edge spec- troscopy up to now.
Fig. 14: Reduced level diagram of W I with observed transitions (in black).
For this line these values are shown in figure 15 for the ground state (TW) as a function of the popu- lation temperature. They coincide quite well with the several (experimental) ones used so far.
Fig. 15: Experimental S/XB values for the 400.8 nm line of neutral tungsten.
Calculated values are shown in red.
2.B. Melting and melt layer propagation
Bulk tungsten is the most promising candidate to replace carbon fibre composites on the divertor targets of ITER. It will also be used during the full metal wall operation of JET. Low sputter erosion and low fuel retention are the most attractive features of tungsten, whereas the major problem is related to melting and melt-layer loss under high-power transient events. Melt layer motion would strongly enhance the local erosion of tiles eventually leading to surface irregularities and formation of tungsten dust or larger droplets.
Fig. 16: Test limiters: (a) graphite roof-like limiter with a tungsten plate, (b) tungsten macro-brush with a castellated structure, (c) tungsten lamellae structure.
The investigation of the melt layer behaviour in a strong magnetic field is, therefore, a necessary step to understand the consequences of erosion phenomena of high-Z refractory metals in fusion devices.
Experiments with three different bulk tungsten test limiters were performed in TEXTOR, as shown in Fig. 16: (a) A thermally insulated solid plate fixed on a graphite roof-like limiter heated up by the plasma to the melting point, (b) a macro-brush of the ITER-relevant castellated structure and (c) a lamellae structure developed for the JET divertor. The main objectives were to determine the dam- age of the metallic surface, the formation of the melt layer and its motion in the magnetic field. The PHEMOBRID-3D and MEMOS-1.5D numerical codes were used to simulate the experiment with the roof-like test limiter. The measured erosion and its numerical simulation are shown in Fig. 17.
The experiments with different types of tungsten limiters in TEXTOR demonstrate that tungsten molten by plasma impact moves rapidly. The motion is perpendicular to magnetic field lines. This motion can be attributed to the thermo-electron emission current and the resulting j×B force. As a
result, tungsten melting can lead to a large material redistribution without ejection of molten mate- rial to the plasma. This effect should be taken into account when assessing the target erosion in ITER.
Fig. 17: Tungsten plate: the measured and modelled toroidally integrated material redistribution along the poloidal direction.
2.C. Other activities dealing with high-Z materials The activities are reported in these sections:
1.F. Investigations of ITER-like castellation structures in the TEXTOR tokamak, 1.H. ERO code and modelling of erosion and deposition, and
4.B. Development of a bulk tungsten divertor module for the highly loaded LB-SRP in JET.
3. Scientific exploitation of new diagnostic systems for the plasma edge 3.A. A new bolometer for JET: operation with partially detached divertor
Partially detached divertor operation is mandatory to reduce strike point power fluxes in long pulse, high power fusion devices. Our understanding of this complex detached state is still incomplete and requires 2D simulations which themselves rely on high quality experimental data to be properly constrained. The radiation distribution is particularly important, but is often poorly resolved in space and magnitude in the divertor region where it is mostly required.
The new bolometer system at JET, the conception, construction and installation of which was en- sured by FZJ staff, provides measurements with significantly improved spatial resolution, allowing the divertor and main chamber radiation fractions to be clearly resolved. L-mode density limit ex- periments have been performed with BT = 2.4 T and Ip = 1.7 MA in ohmic discharges as well as in discharges with additional NBI power of 1.0 to 1.8 MW. With continuous deuterium puffing, a high density, low temperature plasma forms inside the separatrix near the X-point. This is the so-called X-point MARFE (XPM). The XPM, which is the precursor to the ultimate density limit, appears at
about neXPM= 3.0×1019 m-3 at a radiative power fraction γ = Prad/Pheat of about 80%. The neXPMvalue increases with Pheat: neXPM= 3.3×1019 m-3 and neXPM= 3.8×1019 m-3 for auxiliary heating powers of 1.0 MW and 1.8 MW, respectively.
Detachment occurs first at the separatrix and then propagates deeper into the SOL with time. With continuous puffing the outer divertor completely detaches, thereby triggering the X-Point MARFE formation. The general observation is that the plasma is stable if at least one of the divertor legs remains attached or semi-detached. Because the inner leg detaches much earlier than the outer, it is the outer divertor which finally determines the density limit.
The line-integrated intensities measured by the bolometer systems are tomographically inverted by an "anisotropic diffusion tomography model". Fig. 18 shows the tomographic reconstructions of radiation in the divertor region at different ne demonstrating the time evolution of the X-point MARFE formation. The tomographically derived Prad increases with density from 50% to 80% of the total input power just before the XPM formation. increases with density, reaching a maxi- mum value of about 0.6 Pheat just before XPM formation. It drops during the formation of the wall MARFE (WM). The absorbed energy in the divertor derived from thermocouples (ETC = 7.6 MJ) is in good agreement with the total radiated power taking into account the wall load in the di- vertor due to radiation , which is approximately 37% of Prad (ETC ≈ Ein-Erad+ ).
div
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t=55.5s t=55.8s t=56.0s
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ne=3.18×1019 m-3
Fig. 18: Tomographic reconstruction of the total radiation in the divertor region at different ne during the density ramp: 2.2×1019 m-3 and 2.4×1019 m-3: the outer divertor is attached; at 2.56×1019 m-3, the
outer SP starts to detach; at 3.0×1019 m-3: formation of an X-Point MARFE at 80% radiative power fraction.
In contradiction to the earlier observations at the ASDEX-Upgrade and DIII-D tokamaks, it has been observed that the total CD-band emission in the outer divertor increases up to the formation of the WM. Thermal neutrals and also energetic neutrals (generated by charge-exchange) bombard the divertor wall and could provide significant chemical erosion. In this case, the production of the hy-
drocarbons is not localised at the strike point and could take place across the entire divertor wall surface. This is one of the possible explanations for the increase of the line-integrated CD-signal.
3.B. A new bolometer for JET: experiments with an ITER-like magnetic field configuration
A new operational campaign at JET has recently begun, with first attempts at producing high trian- gularity (δ = 0.53), ITER-like plasma configurations. These discharges are ultimately intended to investigate Type I "ELMing" H-modes at high input power and plasma current. They are particu- larly interesting from the divertor physics point of view with regard to the rather short inner divertor leg and the consequent proximity of the X-point to the inner divertor target.
Fig. 19 shows the reconstructed radiation in an inter-ELM period (averaged over 10 ms) and during an ELM (averaged over ELMs) in a discharge with BT = 2.7 T and Ip = 2.5 MA, 14.5 MW NBI and 2.5 MW ICRH power, q95 = 3.6 and n ne GW = 0.87. The ELM frequency was 30 to 35 Hz. Between ELMs, the majority of the radiation is located at the X-point and at the ISP, as well in the outer SOL. At the inner strike point Te is about 43 eV and the inner divertor is attached. Prad (averaged over 10 ms in the inter-ELM period) is 7.75 MW (46% of Pheat) and the wall load in the divertor due to this radiation is roughly 30% of Prad.
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Fig. 19: Total radiation distribution between ELMs (top) and during an ELM (bottom) in a JET discharge with ITER-like configuration.
We find that the radiation during an ELM is – unexpectedly – mainly located in the inner divertor.
The total radiated energy during the ELM is 40 to 50 kJ, which corresponds to 20% of ELM-energy losses (ΔWdiaELM ≈ 240 kJ.)
