Nuclear Fusion Project
Association EURATOM / Forschungszentrum Jülich
A NNUAL P ROGRESS R EPORT 2005
including the contributions of the TEC Partners
ERM/KMS Brussels and FOM Nieuwegein and the IEA Partners
Forschungszentrum Jülich GmbH May 2006
/RPS
A. Introduction and Executive Summary ... 5
B. Institute for Plasma Physics at FZJ (General Description) ... 7
C. Scientific and technological Programme... 9
C.1. Nuclear Fusion and Plasma Research (Summary) ... 9
C.2. Plasma-Wall Interaction ... 23
C.3. Tokamak Physics ... 40
C.4. Plasma Diagnostics... 71
C.5. Contributions to ITER ... 90
C.6. Contributions to Wendelstein 7-X... 103
C.7. Characterisation of Materials and Components under high Heat Loads... 114
C.8. Theory and Modelling... 129
D. Specific Contributions of the Partners within the IEA Implementing Agreement... 152
D.1. Japan... 152
D.2. Canada... 159
D.3. United States of America... 163
E. Summary on results of the main projects in the framework of "Projects for ... 166
enhancing the mutual co-operation between Associations" F. Structure of the FZJ Fusion Programme and Figures... 171
G. List of scientitic Publications, Talks and Posters ... 174
ANNUAL PROGRESS REPORT 2005/ASSOCIATION EURATOM −FZJ A. INTRODUCTION AND EXECUTIVE SUMMARY
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Strategic Relevance and Networking
Research in the field of plasma-wall interaction is of increasing importance for the fusion community in view of the urgent R&D needs for ITER and for other devices, in particular the alternative stellarator concept with long lasting plasma discharges like in LHD or later in Wendelstein 7-X. Plasma-wall in- teraction is seen as a key research area for the realisation of a fusion reactor with high availability and thus low cost of electricity.
The collaboration between the partners of the IEA Implementing Agreement turned out to be an impor- tant element for a further integration and coordination of research on a world wide scale, thus providing the joint use of resources and an added value for all partners. This is in particular important for the field of plasma-wall interaction, since the complexity of related questions requires the employment of a vari- ety of devices ranging from the largest tokamaks to small specialised laboratory experiments. Conse- quently, the work under the Implementing Agreement is not limited only to the use of TEXTOR as a well adapted test-bed for plasma-wall interaction studies, but makes also use of a larger scope of ex- perimental devices (e.g. JET, DIII-D, ASDEX-Upgrade, Tore Supra, JT-60, Alcator C-mod, etc.). Co- ordination of these efforts is provided on the European level via the EU task force “plasma-wall interac- tion” and on the world wide level via the ITPA topic groups. From these activities we also obtain coor- dination among the various tokamak implementing agreements. These implementing agreements have clearly different objectives. Nevertheless, coordination is required where issues overlap or where joint multi-machine experiments are pursued.
The cooperation with neighbouring universities in the field of plasma-wall interaction is further strengthened. After foundation of a Virtual Institute for “ITER-relevant Plasma Boundary Physics” in 2004 – funded by the Helmholtz Association – now also a new Research Training Group on Plasma Physics has been established in 2005. This complements the already existing Collaborative Research Centre on “plasmas far from equilibrium” (e.g. stochastic plasmas). With these activities we further integrate expertise from different disciplines at the universities into the ITER oriented research.
Tasks, Objectives and Main Achievements in 2005
The central facility for the Implementing Agreement is the tokamak TEXTOR, which is operated in collaboration with the partners from the Trilateral Euregio Cluster (Research Centre Jülich, Germany, FOM Rijnhuizen, The Netherlands, and ERM/KMS in Brussels, Belgium). With its highly flexible in- strumentation TEXTOR is oriented towards the investigation of fundamental processes in fusion plas- mas and the pioneering of new concepts. A new versatile tool on TEXTOR is the newly installed Dy- namic Ergodic Divertor (DED), which consists of 16 perturbation coils helically wound around the to- rus on the inboard side allowing to generate magnetic perturbations (stochastic plasmas) and to study their impact on power and particle transport. This new tool, which has been realised with the help of
partners from the Implementing Agreement, widens the scope of research even beyond the specific questions of plasma-wall interaction.
Exploitation of the Dynamic Ergodic Divertor
The DED, which was put into operation in 2003, generates externally applied magnetic perturbation fields which – so far unique – can rotate with a frequency of up to 10 kHz. This opens a new research field. The use of the DED as a powerful experimental tool to study transport, energy and particle ex- haust and plasma stability issues is still at an early stage. The DED is designed for several modes of operation and the corresponding operational scenarios address different aims: while the m/n = 3/1 base mode is the best scenario for the investigation of instabilities in the plasma core, the m/n = 12/4 and the m/n = 6/2 base modes are more suited for the divertor configuration studies and plasma-wall interac- tion. The exploitation of the m/n = 6/2 mode just started at the end of 2005.
The investigation of transport in stochastic plasmas and helical divertor configurations also is of high relevance for stellarators like LHD or Wendelstein 7-X. Some new applications for the DED emerged, unforeseen at the time when the DED was designed: The suppression of pulsed power loads in toka- maks (ELMs) is one of these applications, which is pursued in close collaboration with DIII-D, San Diego, where this method was pioneered. Secondly, the DED has turned out to be a unique tool for the well controlled excitation of tearing modes in the plasma, making the DED in combination with local- ised heating and current drive methods a well suited experimental subsystem for the investigation of plasma stability, a field with high relevance for the operational space of ITER.
Erosion, deposition and tritium retention
Since ITER is on the verge of being built some of the physics questions related to plasma-wall interac- tions need urgent investigation in order to make the best engineering choices. A key question concerns the best choice of materials for plasma facing components in ITER. This is related to research on ero- sion and deposition processes, carbon migration, tritium retention and high-Z plasma-facing compo- nents and mixed material systems. Besides TEXTOR we can rely also on the use of JET, ASDEX- Upgrade, DIII-D, PISCES and others for the experiments needed.
In this context the new ITER-like wall project on JET plays a crucial role. FZJ is working on the design of a bulk tungsten concept for the highly loaded divertor plates in JET thereby making use of the avail- able expertise on engineering and plasma-wall interaction as well as the experimental facilities TEX- TOR and JUDITH. The electron beam facility JUDITH in Jülich is a well-approved instrument for high heat flux testing also of irradiated or toxic materials such as beryllium.
While the detailed investigation of high-Z material behaviour for plasma facing components is of in- creasing importance for the Implementing Agreement, the research on carbon erosion, migration and deposition, as well as the development of methods to diagnose and remove deposited layers is still a major activity within the Implementing Agreement. The ERO-code developed and validated on TEX- TOR provides a key model for the prediction of the amount of tritium retention in ITER.
B. INSTITUTE FOR PLASMA PHYSICS AT FZJ(GENERAL DESCRIPTION)
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The institute 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 making available to mankind a new and practically inexhaustible energy source with favourable safety and environ- mental characteristics.
International fusion research has proved with its experimental facilities that today the principles for the ignition of the fusion fire are known. Now it must be shown that also an economic and continu- ous operation is possible for a large scale power plant. An important step into this direction is the construction of the 500 Megawatt experimental reactor ITER with a tenfold power amplification and a burn duration of approximately eight minutes per plasma pulse. ITER will be realised within a world-wide co-operation at Cadarache in France. The ITER tokamak experiment together with re- sults from the accompanying research programme (materials development, fusion technology, ad- vanced plasma physics) will be decisive for the design of the first fusion power plant DEMO.
The research programme of the institute 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.
The EURATOM-associated fusion laboratories in the Euregio (Institute for Plasma Physics at Re- search Centre Jülich [D], FOM-Institute of Plasma Physics Rijnhuizen [NL] and Laboratoire de Physique de Plasma of the ERM/KMS Brussels [B]) have founded the Trilateral Euregio Cluster (TEC) in order to bundle resources and to favourably bring together different and supplementing expertises. In particular, the TEC performs a common research programme at the TEXTOR toka- mak at Jülich, but also acts as an applicant for certain work packages for ITER (diagnostics port plug). TEC also offers an important point of attraction for the universities in the region. The insti- tute furthermore co-operates with Japan, the USA and Canada in the context of an IEA Implement- ing Agreement.
On the national level the Helmholtz centres Max-Planck-Institute of Plasma Physics Garching, Re- search Centre Karlsruhe and Research Centre Jülich have joined forces within the "Entwicklungs- gemeinschaft Kernfusion" in order to co-ordinate their work. Within Research Centre Jülich, all fusion-relevant work is co-ordinated by the Nuclear Fusion Project (KFS).
Continuous operation of a fusion reactor requires a sufficient life time of the wall components under heavy load as well as the control of stationary plasma confinement under all conditions. Addressing these questions, TEXTOR will in the coming years contribute with its pioneer experiment Dynamic
Ergodic Divertor (DED) and with unique experimental possibilities concerning plasma-wall interac- tion.
The DED generates externally applied magnetic perturbation fields which – so far unique – can rotate with a frequency of up to 10 kHz. This opens a new research field, in particular for the study of transport in stochastic plasmas and helical divertor configurations which are also of high rele- vance for stellarators like Wendelstein 7-X. Some new applications for the DED emerged, unfore- seen at the time when the DED was designed: The suppression of pulsed power loads in tokamaks (ELMs) is one of these applications, which is pursued in close collaboration with DIII-D, San Diego. Secondly, the DED has turned out to be a unique tool for the well controlled excitation of tearing modes in the plasma, making the DED in combination with localised heating and current drive methods a well suited experimental subsystem for the investigation plasma stability, a field with high relevance for the operational space of ITER.
The experiments on plasma-wall interaction at TEXTOR serve as detailed studies of fundamental processes and are meant to supply crucial contributions in the context of a European Task Force addressing the design of ITER. Concerning materials aspects for wall components, a close co- operation is maintained with the Institute for Materials and Processes in Energy Systems at FZJ (IWV-2) and the material-oriented investigations being accomplished there. For strengthening plasma-wall interaction research within the TEC, the new stationary high-flux linear plasma device MAGNUM-PSI is presently being built at the FOM institute in the Netherlands.
Besides TEXTOR, experimental facilities are increasingly used also outside Jülich. Predominantly, the JET tokamak at Culham/United Kingdom is employed for this purpose within the framework of EFDA. There, experimental campaigns are conducted under the leadership of and with contribu- tions from FZJ scientists. FZJ will also participate in significant enhancements of the JET experi- ment during the next years, e.g. in the development of components for the ITER-like wall in JET.
This work relies on a close cooperation of the Jülich institutes IPP, IWV-2 and Central Institute for Technology (ZAT).
The European fusion associations will have to supply their contributions to the planning and build- ing of ITER in accordance with their special expertise. The Institute for Plasma Physics in this con- text aims at taking over task packages from the areas of plasma diagnostics, plasma facing compo- nents and support within TEC to plasma heating.
The construction of the stellarator Wendelstein 7-X in Greifswald is strongly supported by FZJ.
Comprehensive work packages for the construction Wendelstein 7-X have been taken over, the largest of which is the design and fabrication of the superconducting bus-system led by the Institute for Plasma Physics with contributions from ZAT. The institute also develops diagnostic systems and aims at exploring the device in the field of plasma-wall interaction when it becomes operative.
The institute has close links across national boundaries to the neighbouring universities in Germany, Belgium and The Netherlands. In particular the strategic co-operations in plasma physics with HHU-Düsseldorf and RU-Bochum are based on a Research Training Group (Graduiertenkolleg GK 1203), a Collaborative Research Centre (Sonderforschungsbereich SFB 591) and a Virtual Institute.
C. SCIENTIFIC AND TECHNOLOGICAL PROGRAMME
C.1. NUCLEAR FUSION AND PLASMA RESEARCH (SUMMARY)
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With ITER, the globally coordinated nuclear fusion research will achieve the realisation of a first burning fusion plasma with 500 MW power output, eight minutes burning time and tenfold power gain. The ITER tokamak experiment together with results from the accompanying research pro- gramme (materials development, fusion technology, advanced plasma physics) will be decisive for the design of the first fusion power plant DEMO. The stellarator concept is regarded as an attractive alternative candidate to the tokamak due to its specific potential for continuous operation. The op- timised Wendelstein 7-X stellarator in Greifswald, which is currently under construction, will serve to explore the basic suitability of this concept.
The research programme of the Helmholtz Association is geared to the strategy of the European fu- sion research programme, in which the realisation of ITER, ITER-supporting research and the de- velopment of alternative concepts play a central role. Involved in this programme are the Helmholtz centres Max Planck Institute of Plasma Physics (IPP), Research Centre Karlsruhe (FZK) and Re- search Centre Jülich (FZJ) addressing the programme topics ITER, fusion technology, tokamak physics and stellarator research.
ITER
The activities for ITER developments at FZJ comprise the fields of first-wall materials, plasma-wall interaction, diagnostics, heating and current drive.
A major aim of R&D activities in materials research at FZJ is to develop and fabricate new materi- als and to characterise and test them under simulated load conditions (thermal loads, neutron irra- diation). For this purpose the electron beam facility JUDITH-1 is a well-approved instrument for high heat flux testing also of irradiated or toxic materials such as beryllium.
Since ITER is on the verge of being built some of the physics questions related to plasma-wall in- teractions need urgent investigation in order to make the best engineering choices. For this the new ITER-like wall project on JET plays a crucial role. FZJ is working on the design of a bulk tungsten concept for the highly loaded divertor plates in JET. The crucial question which plasma facing components should be made of low-Z and which of high-Z materials is addressed by extensively studying the erosion and deposition behaviour of graphites and carbon-fibre composites, the proper- ties of highly loaded tungsten components, the behaviour of mixed systems and the development of diagnostics.
Fusion technology
FZJ concentrates on the characterisation and test of the thermo-mechanical properties of materials and components under fusion relevant load conditions (reactor specific thermal loads and neutron fluences). With the installation of the new electron beam test facility JUDITH-2, the parameter range of high heat flux simulations is extended and additional testing capacity becomes available.
FZJ is one of 38 partners throughout Europe having joined together in the Integrated EU Project
"ExtreMat" aiming at the creation of new multifunctional materials. FZJ is contributing by provid- ing knowledge on FEM analysis as well as on thermo-shock and thermal-fatigue testing of materials being exposed to high heat loading conditions.
Tokamak physics
Apart from the existing reference scenario for ITER, new modes of operation are also being devel- oped with the aim of enabling longer discharges with higher fusion yield. The major contribution of FZJ consists of investigating the influence of external distortion fields imposed by the Dynamic Er- godic Divertor (DED). This is done on the TEXTOR tokamak in the framework of the Trilateral Euregio Cluster of the three EURATOM-Associations FOM-Rijnhuizen, LPP-ERM/KMS Brussels, and FZJ. Also JET in Culham, ASDEX-Upgrade in Garching, DIII-D in San Diego and Tore Supra in Cadarache are used complementarily, depending on their specific suitability. The major contribu- tions from FZJ to the topic of tokamak physics originate from the areas of confinement and stabil- ity, in particular with the DED, and in plasma-wall interaction.
The DED, which was put into operation in 2003, generates externally applied magnetic perturbation fields which – so far being unique – can rotate with a frequency of up to 10 kHz. This opens a new research field. The use of the DED as a powerful experimental tool to study transport, energy and particle exhaust as well as plasma stability issues already provided a number of remarkable results.
However, the exploitation of the DED is still at an early stage.
Plasma-wall interaction research at FZJ has aligned with the goals of the European Task-Force on PWI. The central motivation is the development of a viable solution to the questions of the ITER first wall and divertor materials. Thus, the research deals with the main fields erosion and deposi- tion, carbon migration, tritium retention, high-Z plasma-facing components and mixed materials.
We can rely on a large span of fusion facilities, i.e. beside TEXTOR also JET, ASDEX-Upgrade, DIII-D, PISCES and others. The IEA-partners (Japan, USA, Canada) are closely linked to this re- search programme as well.
Stellarator Research
According to its existing technical expertise, Research Centre Jülich has taken over comprehensive work packages for the construction of the Wendelstein 7-X stellarator. This includes above all engi- neering work for the design and fabrication of the superconducting bus-system, the bus support structures and joints, stress analysis of support structures and the development of the welding tech- nology of the pressure supports between the main coils. Test of production with dummy conductors has been finished and series production of the bus conductors and the joints will start soon. The de- velopment of diagnostics for Wendelstein 7-X is progressing.
A – Objectives and embedding in the research area
With ITER the globally coordinated nuclear fusion research will achieve the realisation of a first burning fusion plasma with 500 MW power output, eight minutes burning time and tenfold power gain. ITER is based on the tokamak principle, the as yet most advanced concept for the confinement of a hot fusion plasma. The ITER results together with results from the accompanying research pro- gramme (materials development, fusion technology, advanced plasma physics) will be decisive for the design of the first demonstration power plant DEMO – delivering electricity into the grid on a Gigawatt scale in about 30 years. The stellarator concept is regarded as an attractive alternative candidate for a future fusion reactor due to its specific potential for continuous operation. The opti- mised Wendelstein 7-X stellarator in Greifswald, which is currently under construction, will serve to explore the basic suitability of this concept.
B – Programme structure
Within the Helmholtz Association, the Max Planck Institute of Plasma Physics (IPP), Research Centre Karlsruhe (FZK) and Research Centre Jülich (FZJ) are involved in fusion research. All three centres are associated with the European fusion programme. Complementary to the programme topic ITER, which is concerned with the technical construction of this project, the programme topic tokamak physics deals with the physical fundamentals for the realisation of a burning fusion plasma and corresponding concept improvements. Whereas FZK focuses on the programme topics ITER and fusion technology, tokamak physics is exclusively dealt with at IPP and FZJ. In the stel- larator programme topic, the joint construction of Wendelstein 7-X in Greifswald plays a central role. Jülich also contributes to the fusion technology programme topic by the qualification of highly loaded wall materials.
C – Programme results ITER
The experimental facilities of international fusion research have demonstrated the physical princi- ples for igniting the fusion fire. It must now be shown that for an economically feasible power plant a continuous operation is possible. An important step into this direction is the 500 Megawatt ex- perimental reactor ITER constructed in global cooperation with a tenfold power gain and a burning time of about eight minutes per plasma pulse at a site in Cadarache, France.
The research programme of FZJ is geared to the strategy of the European research programme (As- sociation EURATOM-FZJ and European Fusion Development Agreement, EFDA), in which the re- alisation of ITER and ITER-supporting research play a central role. Actual work for ITER com- prises the fields of a) first-wall and divertor materials, b) plasma-wall interaction, c) diagnostics and d) heating and current drive.
First wall and divertor materials
The safe operation of next step tokamaks and stellarators strongly depends on reliable wall compo- nents and in particular with respect to the thermal and mechanical properties of the materials used for plasma facing components. In FZJ a major aim of R&D activity in this field is to develop and fabricate new materials and to characterise and test them under simulated load conditions (ITER
relevant thermal loads, neutron fluences). For this purpose the electron beam facility JUDITH-1 is a well-approved instrument for high heat flux testing. Most of the work is done in the form of tech- nology tasks or physics tasks under EFDA and in cooperation with other fusion laboratories and in- dustry. The work in 2005 addressed the following specific issues:
Manufacturing of plasma facing components
− The casting process of copper onto carbon fibre composites (CFCs) has been studied and opti- mised in collaboration with Politecnico di Torino. Small divertor test modules have been pro- duced and will be tested in JUDITH on their thermal performance under steady state and cyclic loading conditions.
− For the joining of tungsten and copper a new method, namely explosive bonding, is introduced in cooperation with TNO, Rijswijk, the Netherlands, and corresponding studies are initiated.
− For the ITER-like wall project in JET extensive brazing studies for W/CFC joints have been performed on small-scale samples to optimise the brazing parameters.
Material performance during ITER specific thermal loads
− Many specific projects are perused based on manufacturing processes developed in industries:
a) Mechanical and thermo-physical characterisation of carbon and tungsten based materials (SNECMA, Dunlop, Plansee AG), b) Testing of Actively Cooled Divertor Mock-Ups with CFC Armour (ENEA/Ansaldo, Plansee AG and Ansaldo Ricerche) and c) Testing of Actively Cooled First Wall Mock-Ups with monolithic beryllium armour (NNC) and plasma sprayed beryllium coatings (Los Alamos National Laboratories).
− The thermo mechanical behaviour of different bulk beryllium grades and beryllium rich coat- ings on ATJ-graphite under transient thermal loads for pulse durations in the millisecond range has been investigated in JUDITH. Major objective of these experiments are the quantification of the threshold values for cracking and melt layer formation.
− Thermal heat load tests with beryllium-coated specimens and a number of different metallic ma- terials have been evaluated for their suitability to mimic the performance of thin beryllium coat- ings.
− Transient heat load tests with CFC and tungsten targets have been initiated in the plasma gun facility at the Troitsk Institute for Innovation and Fusion Research to study the material re- sponse under ELM-like thermal loads.
− Experimental simulation and modelling (in cooperation with FZK) of material erosion under in- tense energy deposition in off-normal events. JUDITH and a pulsed laser beam are used to un- derstand the mechanism of microscopic and macroscopic erosion during volumetric and surface heating.
Plasma-wall interaction
The activities of FZJ addressing plasma-wall interaction are geared to the critical issues concerning ITER: erosion and deposition of wall materials and the associated retention of the fuel (tritium), de-
velopment methods for limiting the long-term fuel retention and qualification of high-Z materials as alternative plasma-facing materials.
Since ITER is on the verge of being built some of the physics questions related to plasma-wall in- teractions need urgent investigation in order to make the best engineering choices. For this the new ITER-like wall project on JET plays a crucial role. It is planned to test the wall materials foreseen for ITER. FZJ is working on the design of a bulk tungsten concept for the highly loaded divertor plates in JET.
The crucial question – which plasma-facing components should be made of low-Z and which of high-Z materials (e.g. carbon versus tungsten) – is addressed by extensively studying the erosion and deposition behaviour of graphite, the properties of highly loaded tungsten components and the behaviour of mixed systems.
Several EFDA R&D orders for ITER were executed in this field by which the research in Jülich contributes towards the decision making on the most favourable combination of wall materials for the different extension stages of ITER. There are many synergies with the work being performed under the topic tokamak physics (see below). Specific issues addressed are: a) modelling of tritium retention in ITER, b) test of castellated structures, c) material migration under various tokamak con- ditions, d) development of techniques for carbon layer removal and tritium recovery, e) develop- ment of wall conditioning methods under permanent magnetic fields, and f) investigation of melt behaviour under plasma loads.
Diagnostics
The ongoing diagnostic developments are performed in close relation to the needs of ITER, in par- ticular in order to improve the ITER physics base and to prepare for the construction phase of this new international experiment.
− A CXRS system with ITER-relevant geometry (observation from the top) is now being mounted on TEXTOR. This system will allow the simultaneous measurement of the CXRS emission of different impurities, the beam emission and the motional Stark spectrum. The latter measure- ments make it possible to derive the current density profile, whereas the combination of the first two gives a direct measurement of the impurity profile.
− First direct comparative tests of polycrystalline and single crystals as candidate diagnostic mir- rors for ITER have been performed in TEXTOR with respect to their long term reflectivity un- der load conditions.
− Laser induced desorption from deposited carbon layers is proposed as an in-situ diagnostic method to determine and monitor the degree of tritium retention.
− The dispersion interferometer on TEXTOR is the first test of a new scheme for plasma density measurements. It could be ideal for ITER since it is robust against vibrations and thermal ex- pansion. In a next step of the collaboration with the Budker Institute for Nuclear Physics in No- vosibirsk it is envisaged to implement the real-time calculation of the line-integrated electron density.
− FZJ is also participating in a working group of the ITER team studying the engineering of ITER port plugs and the integration of diagnostic systems therein.
Heating and current drive systems
Within the framework of the Trilateral Euregio Cluster (TEC) collaboration, FZJ contributes to the design of an ITER ICRH-antenna. A reduced scale (1/5) mock-up of the antenna array has been constructed and tested at higher frequencies, yielding good agreement of the measured mutual cou- pling properties and the load resilience with the modelling predictions. TEC participates in the con- struction of the ITER-like ICRH antenna project on JET. First tests of the “conjugate-T” matching scheme planned for the ITER antenna have been performed on the new antenna pair installed on TEXTOR. The TEC is also involved, via the partner institute FOM, in the development of a remote- steerable ECRH launching system for the ITER upper ports.
Fusion technology
In addition to the preparations for ITER, the fusion research programme also aims at developing the technologies needed for a later fusion power plant. The first step after ITER will be the demonstra- tion fusion power plant DEMO. The main focus in fusion technology is on the development of the internal wall components and their structural materials.
FZJ as a partner of other materials research laboratories and industries concentrates on the devel- opment and fabrication of new materials and their characterisation and test under simulated load conditions: high thermal loads and relevant neutron fluences. The electron beam facility JUDITH-1 as a well-approved instrument for high heat flux testing also of irradiated or toxic materials plays a central role. With the installation of the new machine JUDITH-2, the parameter range of high heat flux simulations is extended and additional testing capacity becomes available.
An essential part of the ITER-specific material activities performed at Jülich also has a clear rele- vance to future fusion plants such as DEMO (see report on the programme topic ITER); this ap- plies, in particular, to thermo-mechanical investigations of the high-Z material tungsten and its al- loys, the development of joining layers with graded transitions in the tungsten-copper system, and thermal fatigue and thermal shock testing of actively cooled divertor mock-ups.
FZJ is one of 38 partners throughout Europe having joined forces in the Integrated EU Project "Ex- treMat" aiming at the creation of new multifunctional materials. FZJ is contributing on a work package level by providing knowledge on FEM analysis, thermo-shock and thermal-fatigue testing of materials being exposed to high heat loading conditions.
Tokamak Physics
Apart from the existing reference scenario for ITER, new modes of operation are also being devel- oped with the aim of enabling longer discharges with higher fusion yield. That means: on the one hand, the confinement and stability properties must be improved and, on the other hand, steady- state tokamak operation should become possible. This subject is much more explorative than the al- ready far advanced development of the ITER reference scenario. Fundamental work on magnetic confinement, transport behaviour and stability is combined here.
The major contribution of FZJ consists in investigating the influence of external distortion fields imposed by the Dynamic Ergodic Divertor (DED). This is done in the framework of the Trilateral Euregio Cluster of the three EURATOM-Associations FOM-Rijnhuizen, LPP-ERM/KMS and FZJ.
The Jülich contributions to the programme topic of tokamak physics can be subdivided as follows:
1) tokamak physics of confinement and stability, 2) plasma-wall interaction, 3) plasma diagnostics and 4) theory and modelling.
Tokamak Physics of Confinement and Stability
Presently, the main experimental tools to perform our studies at the tokamak TEXTOR are the DED and the auxiliary heating systems ECRH (electron cyclotron resonance heating), ICRH (ion cyclo- tron resonance heating) and NBI (neutral beam injection). The work at TEXTOR is to a significant extent supplemented by investigations at JET, Abingdon/UK and other fusion devices such as DIII- D, San Diego/USA.
The DED, which was put into operation in 2003, generates externally applied magnetic perturbation fields which – so far being unique – can rotate with a frequency of up to 10 kHz. This opens a new research field. The use of the DED as a powerful experimental tool to study transport, energy and particle exhaust and plasma stability issues is still at an early stage. Thus, the main emphasis of re- search in the field of tokamak physics addressing confinement and stability was put on the exploita- tion of the DED.
The DED is designed for several modes of operation, the corresponding operational scenarios ad- dress various different aims. The experiments have shown that the m/n = 3/1 base mode of the DED is the best scenario for the investigation of tearing mode related topics, while the m/n = 12/4 and the m/n = 6/2 base modes are more suited for divertor configuration studies. The exploitation of the m/n
= 6/2 mode just started at the end of 2005. The main results are obtained on the issues 1) divertor structure, 2) electric fields and rotation at the plasma edge, 3) ELM-mitigation, 4) magneto- hydrodynamics (MHD) and 5) profile shaping.
1) Divertor structure
The magnetic topology of the DED forming a helical divertor has been proven. Based on the com- prehensive edge diagnostics on TEXTOR it has been shown that four helical strike zones are formed on the divertor plate. The strike zones can split at certain values of plasma current and mag- netic field forming a private flux zone in between. The plasma flow to the divertor is characterised by a complex magnetic field topology with islands, ergodic zones and laminar flux tubes. The width and the structure of the plasma flux tubes are of particular interest for the plasma flow pattern to- wards the walls. The ergodic regions are characterised by enhanced perpendicular diffusivity for particles and heat.
The investigation of the physics of a helical divertor is of high relevance also for stellarators, since a similar scheme, the island divertor concept, is employed for stellarators like Wendelstein 7-AS, LHD and also for Wendelstein 7-X. Plasma code developments for such complex 3-dimensional structures can now benefit from the detailed experimental findings on TEXTOR. They are e.g. used to benchmark the EMC3 code, which is a joint development of FZJ and IPP.
2) Electric fields and rotation at the plasma edge
Plasma rotation is of particular interest since it can have strong impact on plasma stability and on the suppression of turbulence, thus having relevance for confinement and the operational space of a tokamak. The impact of external distortion fields generated by the DED, either static or rotating, on plasma rotation was a major issue. Experimental results show that plasma rotation is significantly
influenced by the DED. However, the torque leading to rotation can mainly be related to the radial electric field Er produced by the ergodisation rather than to the rotating perturbation field. It has been shown that the torque induced by the DED is comparable to that generated by neutral beams.
3) ELM-mitigation
Edge Localised Modes (ELMs) can lead to very high transient power loads on plasma facing com- ponents and pose a serious thread for their lifetime in ITER. Therefore, techniques are required to limit these loads. Motivated by recent results from the tokamak DIII-D in San Diego, where the suppression of ELMs (ELM mitigation) was demonstrated with the application of external distor- tion fields, we have developed in 2005 a limiter H-mode scenario with ELMs to investigate the ca- pability of the DED for ELM mitigation. This regime is realised with operation at low toroidal magnetic fields of 1.2 to 1.4 Tesla. Being typical for limiter H-mode plasmas, the improvement of the global energy confinement is very small (up to 15% at most), and the enhanced pressure gradi- ent at the plasma edge is predominantly determined by an increased density gradient.
First experiments have shown a reproducible suppression of ELMs by DED operation. Further stud- ies are planned for 2006. This is done to supplement collaborative experiments on DIII-D. More- over, FZJ is participating in a European group starting to assess the options of applying ergodisation coils at JET for the suppression of ELMs.
4) MHD
The DED has turned out to be a unique tool for the well controlled excitation of tearing modes in the plasma, making the DED – in combination with localised heating and current drive methods – a well suited experimental subsystem for the investigation of plasma stability, a field with high rele- vance for the operational space of ITER. These modes are generated by phase locking to the exter- nal perturbation field, i.e. they are locked in the tokamak frame for static (dc) operation of the DED or they rotate with the same frequency as the applied ac DED field.
With these new possibilities TEXTOR became part of the international tearing mode research ac- tivities closely connected to corresponding experiments and theoretical work (e.g. by IPP). Many is- sues have been addressed, as rotation dependence of mode penetration, relation between plasma beta and mode threshold, detailed studies of the mode excitation threshold as function of the tor- oidal plasma fluid rotation, mode locking and unlocking experiments, active control of mode fre- quency by using the rotating perturbation fields, suppression by ECRH and ECCD of tearing modes purposely induced by the DED, modelling of the penetration process of the DED field and demon- stration of the mass dependence of the frequency of Alfvén-like modes.
5) Profile shaping
Active shaping of plasma profiles by local heating and current drive is a promising way for further improvements of the tokamak concept. In this respect some experiments with localised electron cy- clotron resonance heating and current drive were performed on TEXTOR. In particular, a joint ex- periment with the Russian tokamak T-10 concentrated on switch-on and switch-off effects of off- axis ECRH addressing, the issue of transient regimes of improved confinement and reduced turbu- lence.
Plasma-wall interaction (PWI)
This area is of basic significance for the operation of ITER and, in particular, for the development of a steady-state burning fusion plasma. Suitable wall materials must combine low erosion with low hydrogen retention to obtain a fuel inventory as low as possible. At present, a combination of car- bon, beryllium and tungsten is envisaged for the plasma-facing components in ITER.
PWI research in FZJ has aligned with the goals of the European Task-Force on PWI. The central motivation is the development of a viable solution to the questions of the ITER first wall and diver- tor materials. This policy ensures that ITER-relevant subjects are treated with due priority. Thus, the research deals with the following main fields:
• Erosion and deposition, carbon migration and tritium retention, and
• high-Z plasma-facing components and mixed materials.
We can rely on a large span of fusion facilities, i.e. besides TEXTOR also JET, ASDEX-Upgrade, DIII-D, PISCES and others. Any specific physical investigation is thus carried out on the best suited machine. The IEA-partners (Japan, USA, Canada) are closely linked to this research programme as well. The main results obtained in 2005 are summarized in the following.
Erosion and deposition, carbon migration and tritium retention
− Experiments for the improvement of spectroscopic methods to determine the erosion yield, in particular via the hydrocarbon break-up products C2 or CH, have been performed under the lead of a scientific coordinator from FZJ at JET, TEXTOR and ASDEX-Upgrade.
− Fulcher-band spectroscopy, already successfully applied on TEXTOR, has been transferred to JET as a sensitive method to detect a low minority concentration (H or T) versus a high deute- rium background level. This is important for the understanding of particle fluxes and in particu- lar the retention of tritium.
− The investigations of fuel accumulation, carbon migration and materials mixing in gaps of cas- tellated structures have been continued. The thickness of deposited layers has been determined and also the corresponding fuel accumulation. Efforts for modelling these observations are con- tinuing.
− Studies are ongoing with RF supported DC glow discharges in gas mixtures of O2 and He as a possible technique to remove deposited carbon layers. ICRF discharge conditioning is proposed as an alternative technique for the next generation of super-conducting fusion machines. Inter- machine studies on this topic have been performed on TEXTOR, ASDEX-Upgrade and JET.
High-Z plasma-facing components and mixed materials
− Plasma-facing components made of high-Z materials and systems of mixed materials have been investigated according to different aspects. A brush limiter (castellated surface) which consists of cut tungsten slices on a copper base target has been exposed to the TEXTOR plasma to study the deposition of materials in gaps and to investigate the behaviour of these surfaces at extreme heat loads which lead to melting events. The melt layer flow observed in these cases can be ex- plained by forces due to a combination of thermo-electronic emission and the magnetic field.
Large material redistributions are observed without ejection of molten material into the plasma.
− Also tungsten coatings on graphite were exposed to the TEXTOR plasma and analysed post mortem.
− The method of spectroscopic determination of tungsten release has been improved by new cal- culations on conversion factors S/XB (ionisations per photon) for specific lines and comparison with experimental results from TEXTOR.
Modelling and prediction of tritium retention
Modelling of erosion and deposition with the ERO-code has been further improved by benchmark- ing with well defined experiments. Amounts of 13CH4 are injected and the evolution of deposition patterns is studied. The parallelisation of the ERO-code has been completed, now allowing for a de- crease of calculation time by about a factor of 5 to 10. The coupling with the Monte-Carlo code SDTrimSP now allows to simulate the dynamic change of the surface composition in mixed mate- rial systems. The modelling of experiments with beryllium in the linear plasma device PISCES has been started aiming at an improved understanding of deposition and erosion of mixed Be-C-W lay- ers. The ERO-code provides a key model for the prediction of the amount of tritium retention in ITER.
Plasma diagnostics
The development of methods and techniques for measuring important plasma parameters is an inte- gral part of fusion research and thus an over-arching component of all programme topics. FZJ and its TEC-partners have developed a broad expertise in diagnostics. The new developments are de- scribed here, while diagnostic projects more directly linked to the new machines ITER and Wendel- stein 7-X are noted in the corresponding chapters of these topics. FZJ together with its TEC- partners is a strong candidate to become responsible for the design and construction of one of the instrumented ITER port plugs. Highlights of diagnostic developments for the plasma edge and core to be applied in TEXTOR, JET, LHD and others are given in the following.
Edge diagnostics
− Significant work has been put on the adaptation of diagnostics to the new challenges given by the complex 3-dimensional plasma structure generated by the DED (atomic beams, charge ex- change spectroscopy, thermography, Thomson scattering, probes).
− The spectroscopic methods to detect carbon atoms and molecules have been significantly im- proved. Atomic and molecular lines of hydrocarbon break-up products can be used for a better determination of carbon erosion. The laser absorption spectroscopy technique for the detection of CH4 was further developed and calculations with the code package “ATOM” have been made to improve our data base on atomic data for diagnostics. Based on this expertise FZJ has got a contract to refurbish some spectroscopic systems on JET during the next years.
− Some diagnostic system are developed in cooperation with universities (Bochum, Düsseldorf):
an ellipsometric system for in-situ real-time measurements of erosion-deposition processes and a fast electron detection system.
− A new Nuclear Reaction Analysis (NRA) system for post-mortem analysis of samples having been exposed to TEXTOR is to be installed at the tandem accelerator facility at FZJ/ISG and will be operational in the beginning of 2006.
− A penning spectroscopy system was set up at the stellarator LHD in Japan to measure the partial pressure of neutrals in the main chamber.
− Six new quartz microbalance diagnostic systems (QMB) have been installed in remote areas of the inner and outer divertor on JET to allow for time resolved measurements of the amount, dis- tribution and composition of deposits.
− A fast valve with high gas throughput has been developed and built for JET.
− As a joint development between FZJ and FOM a new edge Thomson scattering system was fur- ther developed; this multi-pulse ruby laser is capable of delivering 3 bursts of 30 pulses each with pulse energy of about 15 J at 5 kHz. The TEXTOR plasma itself is part of the 18 m long laser cavity
Core diagnostics
− An ultra fast camera system for disruption analysis and observation of pellet injection, for the study of edge transport phenomena, and for fast divertor spectroscopy was taken into operation and prepared for integration at TEXTOR.
− The magnetic diagnostics were improved in various ways for a better and more reliable plasma control in TEXTOR.
− Four sets of new soft x-ray cameras viewing the plasma vertically and horizontally will be very useful for the investigation of fast MHD activities on TEXTOR.
− A 10 kHz multi-position Thomson scattering diagnostic was commissioned being capable of measuring electron temperature and density profiles with high spatial resolution.
− Work on a combined Microwave Imaging Reflectometry (MIR) and Electron Cyclotron Emis- sion Imaging (ECEI) system, capable of measuring density and temperature fluctuations, has progressed. The 2D ECE Imaging system on TEXTOR has come into routine operation while the MIR system is still in the commissioning phase.
− A 16-channel tuneable-frequency ECE radiometer has been installed and commissioning was started by the end of the year.
− The development of position sensitive nuclear track detectors for TEXTOR was continued in collaboration of ERM/KMS with the Zoltan Institute of Swierk (Poland).
− As the only present tokamak in the world, TEXTOR holds the capability of measuring the local- ised 1-dimensional distribution function of the confined fast ions using the technique of collec- tive Thomson scattering (CTS). This diagnostic setup has now been upgraded at Risø and was reinstalled at TEXTOR.
− Two new bolometer cameras (KB5) for horizontal and vertical views of the plasma cross section have been installed on JET. The new cameras lead to a substantial improvement of the recon- struction of radiation in the divertor region.
Theory and Modelling
A wide spectrum of theoretical and computational topics in magnetic fusion research has been cov- ered. This ranges from direct numerical support of experiments (3D tokamak magnetics, including island and stochastic layer formation, equilibrium reconstruction for TEXTOR-DED, Wendelstein 7-X, rarefied gas dynamics applied to vacuum system design, ITER) up to first principle theoretical studies of turbulence on small scales and low (drift) frequencies, applied to the ergodised edge plasma layer of TEXTOR-DED.
Major long term code developments at FZJ for the European fusion program, such as the ERO code for local erosion and re-deposition simulation, or the Monte Carlo neutral gas and radiation transfer code EIRENE, have been integrated into the JET and EFDA ITM (integrated tokamak modelling) task forces. The goals are both to standardise the data exchange formats between codes in Europe but also to ensure quality control (code verification and validation). Both these codes have also been adapted and made available to IEA partners and are routinely applied to assist the ITER engineering design. A joint web based project between FZJ and the IAEA atomic data unit has been established to further improve the predictive quality of fusion plasma codes by developing appropriate and standardised atomic and molecular databases.
First 3D edge plasma fluid codes, developed for stellarator applications, such as EMC3-EIRENE (for Wendelstein 7-X), are currently extended and validated using experimental data from the 3D edge plasma experiments on TEXTOR under DED operation. This is a joint effort between FZJ, IPP-Greifswald, ITER and NIFS (LHD, Japan).
New theoretical results on internal transport barriers and the effect of magnetic field ergodization, mainly resulting from first principle anomalous (turbulent) transport studies, are quantified and as- sessed by applying them in radial transport code studies for TEXTOR and JET parameters.
Stellarator Research
The Max Planck Institute of Plasma Physics (IPP) currently builds the Wendelstein 7-X stellarator in Greifswald – with contributions from FZJ and FZK. The stellarator concept is regarded as an at- tractive candidate for a future fusion reactor due to its specific potential for continuous operation.
Wendelstein 7-X is a large stellarator, which has been optimised according to the quasi-symmetry principle. It consists of superconducting coils and is intended to provide plasma discharges of 30 seconds duration at a heating power of 10 MW. The aim is to demonstrate the basic suitability of the chosen concept for magnetic confinement with long pulses.
According to its existing expertise, FZJ has taken over comprehensive work packages for the con- struction of Wendelstein 7-X. This includes above all work for the design and manufacturing of components of the superconducting coils, of the leads and electrical connections, supporting work in welding technology, strength calculations as well as diagnostics development. The major work in FZJ carried out in the year under review concerns engineering and welding technology tasks, while the development concerning two large diagnostic systems represents a smaller amount of work.
Engineering
Superconducting Bus-System: The coils are interconnected by superconducting bus bars. The tech- nical specifications are the basis for the design, construction, qualification, manufacturing and as- sembly of the buses and their appropriate supports.
− A new bus topology was developed to avoid collisions with other parts and to facilitate the as- sembly. To check the geometry of the bent buses and to examine the buses assembly, a 1:1 model was manufactured and assembled.
− For the qualification of the insulation and the fabrication process different samples have been produced and a special examination procedure has been developed.
− For series production of the 125 buses a production line has been installed in Jülich. One of the most critical points to be solved was the handling of the up to 13 m long 3-dimensionally shaped bus bars. Test of production with dummy conductors has been finished and series pro- duction will start soon.
Bus support structure: The design of the support structure is based on different adjustable sub- modules which are able to compensate fabrication tolerances in all directions and to facilitate the assembly on site. Movement and forces acting on the buses are taken into account during several it- erations of support design and stress calculations.
Joints: Approximately 230 low-resistance joints are required for electrical and hydraulic intercon- nections between superconductors at the coil terminals and between five adjacent modules. After design review three joints have been manufactured and tested under pressure. Resistance tests at 4 K are in preparation and after delivering of material the manufacturing of 230 joints including inner clamping parts will be started at FZJ.
Stress analysis: The structural analyses of the numerous options of the narrow support elements have been performed with finite element models. Requirements for the bus-bar supports have been developed. The design changes in the joint structure were modelled.
Welding Tests: FZJ develops the technology for welding the pressure supports between the magnet coils. In the course of the year, altogether 6 welding attempts were executed in order to determine and to optimise the occurring welding distortion. These welding tests guarantee that the necessary weld quality can be achieved on a high level also at the actual experiment. The manufacture of the welds puts high demands on the dexterity of the executing welder. In February 2005 two welders of the IPP-Greifswald were trained in the FZJ.
Diagnostics
VUV spectrometer: In the course of 2005, the construction of the 4-channel VUV spectrometer sys- tem HEXOS (High Efficiency XUV Overview Spectrometer) has been completed together with all peripheral equipment such as mechanical stands, vacuum systems, detectors, cameras and data ac- quisition which are needed to perform the laboratory testing. The spectrometers have been assem- bled and aligned at the Horiba Jobin-Yvon laboratories. The alignment of the HEXOS spectrometer channels was tested by measuring spectra from a pinch discharge (by AIXUV GmbH, Aachen) and
a hollow cathode discharge. The measured line width was found to be in agreement with the ex- pected values.
Hydrogen beam: A high energy hydrogen beam diagnostic injector is developed in cooperation with the Budker Institute of Nuclear Physics (BINP) in Novosibirsk. The beam shall provide an equiva- lent current of more than 5 A at 60 keV energy. During the whole duration of injection (i.e. 10 s) the beam properties the divergence should stay below 0.5º. The R&D work on optimisation of the grid structure for higher current densities and lower beam divergence of the ion optics was contin- ued on TEXTOR. The decision about the actual type of ion source will be made based on the per- formance of an existing prototype at TEXTOR.
D – Further programme development
In the course of this year the legal framework for the realisation of ITER has to be defined. The European Legal Entity (ELE) once founded will be responsible for all procurements to ITER pro- vided by Europe. Thus, the ELE will be the important contractual partner for all fusion associations involved in work for ITER. All R&D in addition to the construction of ITER (JET, physics and gen- eral technology) will remain with the European Fusion Development Agreement (EFDA). The HGF centres involved will have to re-orientate their activities accordingly. A crucial point will be the for- mation of European consortia, which will be charged with certain work packages for ITER. FZJ is presently involved in discussions about participation in consortia on diagnostics (instrumented port plug), plasma facing components and heating. This is supported by a European training programme for the development of human resources in fusion engineering.
The JET experiment, which is prolonged until 2010, will play a central role for the European fusion research.
The construction work for Wendelstein 7-X will go on for the next years. Commissioning and first experiments are expected after 2012. FZJ is aiming at taking a leading role in the use of Wendel- stein 7-X for plasma-wall interaction research.
The recently approved new linear plasma device MAGNUM-PSI will be built at the site of our TEC-partner FOM in Rijnhuizen, The Netherlands. This high flux steady state device will provide a new powerful tool for plasma-wall interaction research, in particular on issues which cannot be covered by pulsed tokamak plasma operation. FZJ will contribute to the construction of the device by designing the target station and will later be partner in exploiting this device.
Strategic cooperation with neighbouring universities is of increasing importance. It appears neces- sary to enforce education in the field of plasma physics and nuclear fusion. In addition, research at the universities has to linked with international fusion research in a much better way.
IPP, FZK and FZJ plan to modify the HGF fusion programme topics in order to streamline them more efficiently to the available expertise in the research centres: 1) ITER, 2) tokamak physics, 3) stellarator research, 4) plasma-wall interaction, 5) theory and modelling and 6) fusion tech- nology.
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 quantification 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.
