Nuclear Fusion Project
Association EURATOM / Forschungszentrum Jülich
A NNUAL P ROGRESS R EPORT 2004
including the contributions of the TEC Partners
ERM/KMS Brussels and FOM Nieuwegein and the IEA Partners
Forschungszentrum Jülich GmbH November 2005
/RPS
CO N T E N T S PA G E
A. Introduction ... 5
A.1. Nuclear Fusion ... 7
B. General Programme on TEXTOR ... 17
B.1. Main Topic I − Plasma Wall Interaction ... 17
B.2. Main Topic II − Confinement and Transport ... 26
B.3. Main Topic III − Perturbation Field Effects and MHD ... 37
B.4. Main Topic IV − Theory and Modelling ... 48
B.5. Plasma Diagnostics ... 61
B.6. Contributions to ITER ... 74
B.7. Contributions to Wendelstein 7-X ... 87
B.8. Characterization of Materials and Components for Plasma/Wall Interaction ... 92
C. Technology Programme... 98
C.1. Characterization of Materials and Components for Plasma/Wall Interaction ... 98
C.2. Oxidation Resistance of innovative C-based Materials ... 103
C.3. Mechanical Properties of Fusion Materials ... 116
D. Partners of the IEA TEXTOR Implementing Agreement ... 117
D.1. Japan ... 117
D.2. Canada ... 125
D.3. United States of America... 129
E. Summary on results of the main projects in the framework of "Projects for ... 131
enhancing the mutual co-operation between Associations" F. Structure of the Fusion Programme and Related Figures ... 135
G. Scientific Publications, Talks and Posters ... 138
ANNUAL PROGRESS REPORT 2004/ASSOCIATION EURATOM −FZJ A. INTRODUCTION
räêáÅÜ=p~ããI=oçÄÉêí=tçäÑ=EfmmF u.samm@fz-juelich.de
r.wolf@fz-juelich.de
The institute is part of the international nuclear fusion research, which pursues the long-term goal of realizing on earth the method of power generation employed by the sun, thus making available a new and practically inexhaustible energy source with favourable safety and environmental characteristics to mankind.
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 continuous op- eration 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. The realization of ITER is planned within a world- wide co-operation. The results of ITER will be crucial for the design of the first demonstration power station DEMO.
The research programme of the institute is oriented towards the strategy of the European research pro- gramme (Association EURATOM-FZJ and European Fusion Development Agreement EFDA), where the realization of ITER and the research in support for ITER play a central role.
The EURATOM-associated fusion laboratories in the Euregio (Institute for Plasma Physics at Research Centre Juelich [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 tokamak at Jülich, but also acts as an applicant for certain work packages for ITER (diagnostics port plug) and plans the construction of a new stationary linear plasma device at the location of the FOM institute in the Netherlands. TEC also offers an important point of attraction for the universities in the region. The institute furthermore co- operates with Japan, the USA and Canada in the context of an IEA Implementing Agreement.
On the national level the Helmholtz centres Max-Planck-Institute of Plasma Physics Garching, Re- search Centre Karlsruhe and Research Centre Juelich have joined forces within the "Entwicklungsge- meinschaft Kernfusion" in order to co-ordinate their work. Within Research Centre Juelich, all fusion- relevant work is co-ordinated by the Nuclear Fusion Project (KFS).
Continuous operation of a fusion reactor requires a sufficient life span 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 interaction.
The DED allows to study fundamental options and mechanisms by means of additional external mag-
netic perturbative fields in the plasma boundary region (stochastic plasmas) and also by means of stabi- lization effects further inside the plasma.
The experiments on plasma-wall interaction at TEXTOR serve as detailed studies of fundamental proc- esses 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 main- tained with the Institute for Materials and Processes in Energy Systems (IWV-2) and the material- oriented investigations being accomplished there.
Besides TEXTOR, experimental facilities are increasingly used also outside Juelich. 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 contributions from FZJ scientists.
The European fusion associations will have to supply their contributions to the planning and building of ITER in accordance with their special expertise. The Institute for Plasma Physics in this context aims at taking over task packages from the areas of plasma diagnostics and plasma heating.
Because of its inherently stationary plasma operation, the stellarator is considered as the most promis- ing alternative to the tokamak. With the stellarator Wendelstein 7-X at Greifswald – expected to be- come operational in 2010 – Germany will possess a world-wide leading experiment in this research area. The Institute for Plasma Physics contributes to the construction of Wendelstein 7-X with electro- technical tasks and with the development and supply of diagnostic systems. The institute will also par- ticipate in the scientific use of the new stellarator at a later stage.
ANNUAL PROGRESS REPORT 2004/ASSOCIATION EURATOM − FZJ A. INTRODUCTION
A.1. NUCLEAR FUSION
räêáÅÜ=p~ããI=oçÄÉêí=tçäÑ=EfmmF u.samm@fz-juelich.de
r.wolf@fz-juelich.de
Summary
The medium-term goal of globally coordinated nuclear fusion research is the realization of a burn- ing plasma with 500 MW fusion power, eight minutes burning time and tenfold power gain. This is to be achieved in global cooperation by the construction of the ITER project which will be decided upon before long. The ITER results will be decisive for the design of the first demonstration power plant DEMO.
Apart from that, the stellarator concept is also regarded as an attractive candidate for a future fusion reactor due to its specific potential for continuous operation. The optimized Wendelstein 7-X stel- larator in Greifswald, which is currently under construction, will serve to explore the basic suitabil- ity of this concept.
The research programme of the Helmholtz Association is geared to the strategy of the European fusion research programme, in which the realization of ITER, ITER-supporting research and the development of alternative concepts play a central role. Involved in this programme are the Helm- holtz centres Max Planck Institute of Plasma Physics (IPP), Research Centre Karlsruhe (FZK) and Research Centre Jülich (FZJ) with the programme topics ITER, fusion technology, tokamak physics and stellarator research.
ITER
The activities for ITER at Research Centre Jülich comprise the fields of first-wall materials, plasma-wall interaction, diagnostics, heating and current drive.
Modules of components subjected to high thermal fluxes were tested in the JUDITH electron beam facility with respect to their heat removal performance and their thermal fatigue in conjunction with neutron irradiation tests with ITER-specific neutron fluences of up to 1 dpa at temperatures in the range of 200 – 700 °C. A new electron beam test facility, JUDITH 2, was set up for the simulation of extremely short transient load scenarios. The new system is capable of also performing thermal cycling tests on large components.
With its work on plasma-wall interaction Jülich contributes towards making a decision on the most favourable combination of wall materials for the different extension stages of ITER. Various activi- ties for ITER diagnostics projects are carried out at FZJ using TEXTOR as the test environment.
Jülich contributes, in particular, to the topics of plasma heating and current drive in ITER via the two TEC partners FOM and ERM/KMS.
Fusion technology
Fusion technology is a cross-sectional topic for Jülich, to which a contribution is made, in particu- lar, by the qualification of highly stressed wall materials – with relevance to ITER, Wendelstein 7- X and DEMO.
Tokamak physics
The focus of fusion research at Research Centre Jülich is on the topics of a) stochastic plasmas with the Dynamic Ergodic Divertor and b) plasma-wall interaction. This programme contributes to im- proving the tokamak principle, consolidating and extending the physical fundamentals for the de- sign and operation of ITER and preparing the scientific use of Wendelstein 7-X.
For the implementation of the programme, the TEXTOR tokamak jointly operated within the framework of the Trilateral Euregio Cluster (TEC) is available at FZJ. For the programmatic topics, the largest tokamak worldwide, JET in Culham/Great Britain, ASDEX Upgrade in Garching, DIII- D in San Diego/USA and Tore Supra in Cadarache/France are also used complementarily, depend- ing on their specific suitability.
The Dynamic Ergodic Divertor (DED) put into operation at TEXTOR in 2003, with which the en- ergy and particle exhaust from the plasma is to be improved by magnetic perturbation fields, has demonstrated its basic functionability and has, moreover, opened up access to many new issues.
This includes, in particular, the influence of the magnetic perturbation field on the confinement and stability of the plasma.
In the field of plasma-wall interaction, Jülich plays a leading role in the coordination of European research activities. In accordance with the requirements for a suitable combination of wall materials with low erosion and low hydrogen retention along with heat load control, current work at Jülich is concerned with the chemical erosion of graphite, with its migration behaviour, the qualification of decomposition methods for deposited carbon, research into alternative materials and the develop- ment of new measuring techniques.
Stellarator
According to its existing technical expertise, Research Centre Jülich has taken over comprehensive work packages for the construction of the Wendelstein7-X stellarator. This includes above all work for the design and fabrication of components of the superconducting coils (bus system), supporting work in welding technology, strength calculations, materials issues and diagnostics development.
The fabrication of the bus system and of all the 250 low-resistance joints will start at Jülich in 2005.
The JUDITH electron beam facility was used to test the properties of actively cooled components for the divertor targets of Wendelstein 7-X at high heat load. The development of diagnostics for this stellarator was further advanced with TEXTOR as the test bed.
Objectives and embedding in the research area
The medium-term goal of globally coordinated nuclear fusion research is the realization of a burn- ing plasma with 500 MW fusion power, eight minutes burning time and tenfold power gain. This is to be achieved by the construction of the ITER project which will be decided upon before long.
ITER is based on the tokamak principle, the as yet most advanced concept for the confinement of a hot fusion plasma.
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 realization of this project, the programme topic tokamak physics deals with the physical fundamentals for the realization of a burning fusion plasma. Whereas FZK focuses on the programme topics ITER and fusion technology, tokamak physics is exclusively dealt with at IPP and FZJ. In the stellarator programme topic, the joint con- struction of the Wendelstein 7-X stellarator in Greifswald plays a central role. Jülich contributes, in particular, to the fusion technology programme topic by the qualification of highly stressed wall materials with relevance to ITER, Wendelstein 7-X and DEMO.
Programme results ITER
International fusion research has demonstrated with its experimental facilities that it knows the physical principles for igniting the fusion fire. It must now be shown that economical continuous operation on a power plant scale is possible. An important step in this direction is the planned 500 megawatt ITER experimental reactor to be constructed in global cooperation with a tenfold power gain and a burning time of about eight minutes per plasma pulse. The ITER results will be decisive for the design of the first demonstration power plant DEMO.
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 realization of ITER and ITER-supporting research play a central role. Concrete work for ITER comprises the fields of a) first-wall materials, b) plasma-wall interaction, c) diagnostics and d) heat- ing and current drive.
Plasma-interactive materials and components for high heat loads
Within the framework of materials development for thermally highly stressed components of ITER, the industry and other research institutions have manufactured a variety of miniaturized wall mod- ules covering a wide range with respect to design, the connection technologies used and the choice of the plasma-interactive material (beryllium, CFC, tungsten). These components were tested in the JUDITH electron beam facility; besides determining the heat removal performance, particular atten- tion was paid here to damage caused by thermal fatigue.
Modules with CFC or tungsten reinforcement envisaged for the divertor of ITER have sustained exposure to thermal fluxes of up to 25 MW per square metre under cyclic load (n = 1000) without damage in these experiments and thus meet the specifications required by ITER. Components based on the monoblock design proved to be more damage-tolerant in these experiments compared to the flat-tile concept. In addition to the two materials to be used for the divertor (carbon and tungsten), a variety of beryllium-reinforced modules for the first wall of ITER were also successfully tested.
Neutron irradiation tests were performed on actively cooled heat sinks of the tungsten-copper and CFC-copper material systems (monoblock and flat-tile concept), applying ITER-specific neutron fluences of up to 1 dpa (displacement per atom) at temperatures in the range of 200 – 700 °C. In subsequent heat flux simulation tests, a distinct temperature increase was observed for the CFC components, which is primarily attributable to a decrease in thermal conductivity of the carbon ma- terial. For components with tungsten reinforcement the decrease in thermal conductivity is much
less important. The limit loads attainable in thermal cycling tests were 15 MW per square metre for CFC components and 18 MW per square metre for tungsten modules. An extensive irradiation pro- gramme (0.6 dpa) was initiated for beryllium-reinforced wall modules. The post-examination of the irradiated components will take place in the JUDITH electron beam facility.
For the simulation of extremely short transient load scenarios (ELMs with pulse durations in the sub-ms range) and for the investigation of synergistic effects (interaction of thermal shock and thermal fatigue loads occurring at the same time) a new electron beam test facility, JUDITH 2, was planned and set up in the "Hot Materials Laboratory" (HML) specifically prepared for this purpose.
In addition to transient material and component tests, the new system is also capable of performing thermal cycling tests on large components with a loadable area of up to 500 mm x 500 mm. A vari- ety of measuring systems are available for diagnostics.
Plasma-wall interaction
The activities of Research Centre Jülich on plasma-wall interaction are geared to the critical issues concerning ITER: erosion of wall material and the associated retention of the fuel (tritium), devel- opment of in-situ control and decomposition methods for limiting long-term fuel retention, qualifi- cation of high-Z materials as alternative plasma-interactive material and limitation of peak loads by transient heat pulses. With these activities Jülich contributes towards making a decision on the most favourable combination of wall materials for the different extension stages of ITER (see also To- kamak physics). Several concrete R&D orders for ITER were executed in the fields of carbon mi- gration, decomposition techniques for deposited carbon, measuring techniques for the determination of tritium capture and further development of numerical codes.
Diagnostics
Based on EFDA contracts and on the use of TEXTOR as a test environment, activities for two ITER diagnostics projects are currently being carried out at FZJ:
a) The design of a six-channel VUV spectrometer system for ITER, which covers the wavelength range from 2.3 nm to 160 nm (divided into six subsections with overlap) has been completed.
The frequency resolution of the system permits the observation of relevant plasma impurities with the precision required in ITER.
b) The design of the ITER-CXRS diagnostics for charge exchange, which is based on the utiliza- tion of a particular neutral particle beam, is being further developed and tested in a joint effort within TEC and together with other partners. The detailed optical design of the observation periscope, the fibre optics and the spectrometers was improved by ray tracing methods and the measurement options were analysed on the basis of modelling the charge-exchange processes and the beam emission spectra. The system is being optimized to permit a simultaneous deter- mination of ion density, temperature, plasma rotation and magnetic fields over a wide radial range.
Furthermore, design work for a poloidal polarimetry system and a wide-angle thermography system was carried out within the framework of the TEC collaboration.
Within the framework of TEC, FZJ plans to take over the coordination for the development and manufacture of a complete diagnostic unit (a so-called "port plug") for ITER. One aspect is here to facilitate access to the scientific programme of ITER.
Heating and current drive systems
The plasma heating in ITER serves, on the one hand, for reaching the temperatures needed for the onset of the burning process and, on the other hand, for subsequently controlling the burning fusion plasma. This includes the suppression of instabilities and also the current drive with the aim of re- placing, at least in part, the inductive plasma current.
The two TEC partners FOM and ERM/KMS are mainly active in this field. The Dutch FOM insti- tute operates the microwave heating (resonant heating of the plasma electrons) in TEXTOR, which can also be used for current drive. It has been possible with this heating method to stabilize tearing modes again that were excited in a controlled manner by the DED. At present, a combined heating and measuring method is being worked on, which would allow the detection and subsequent stabili- zation of the mode using a single instrument. TEC is involved in the development of a remotely controllable ECRH antenna for one of the upper ITER observation windows.
The Belgian TEC partners operate the radiofrequency heating (resonant heating of plasma ions) in TEXTOR. Similar to the antennas envisaged for ITER and also to be tested at JET, a new antenna was installed in TEXTOR in order to investigate the coupling between antenna and plasma. This antenna will be used for the first time in 2005 to enable an extrapolation for the corresponding ITER heating on the basis of JET experiments.
Fusion technology
In addition to the preparations for ITER, the fusion research programme aims at developing the technologies needed for a fusion power plant. The first step after ITER will be the DEMO demon- stration fusion power plant. The main focus in fusion technology is on the development of the inter- nal wall components and their structural materials. An essential part of the ITER-specific material activities performed at Jülich has also a clear relevance to future fusion plants such as DEMO (see report on the programme topic ITER); this applies, in particular, to thermomechanical investigations of the high-Z material tungsten and its alloys, to the development of joining layers with graded tran- sitions in the tungsten-copper system and to the construction of the new JUDITH 2 test facility.
In the year under review, moreover, investigations were performed on the influence of implanted helium on the properties of various materials (iron, EUROFER97, copper, tungsten). In particular, recovery effects with respect to embrittlement by annealing the specimens were studied.
Tokamak physics
The Jülich contributions to the programme topic of tokamak physics can be subdivided as follows:
i) improving the tokamak principle, (ii) plasma-wall interaction, (iii) consolidating and extending the physical fundamentals for the design and operation of ITER, (iv) developing plasma diagnostics and (v) theoretical tokamak physics. These component parts of the programme topic first of all aim at the construction and operation of ITER, but they also contain follow-on aspects of the utilization of fusion as an energy source, which requires the steady-state operation of a fusion power plant.
For the implementation of the programme, the TEXTOR tokamak is available at Jülich, for the effi- cient use of which the Jülich researchers have joined forces in the Trilateral Euregio Cluster (TEC) with the FOM Institute for Plasma Physics Rijnhuizen in the Netherlands and the Laboratoire de Physique des Plasmas of ERM/KMS in Belgium. This cooperation meanwhile also extends to the joint development and design of diagnostics and plasma heatings for ITER. With its flexible in- strumentation TEXTOR is oriented towards investigating fundamental processes of hot fusion plasmas. With the increasing integration of the European and global fusion programme, experi-
ments are not only performed at TEXTOR, but also at JET, ASDEX Upgrade (IPP) and DIII-D (General Atomics, USA). A major organizational aid is provided by the "International Tokamak Physics Activity" (ITPA), which coordinates experiments for dealing with important issues (joint experiments) at the tokamaks available worldwide.
In 2003, after an installation phase of more than two years, the Dynamic Ergodic Divertor (DED) was put into operation at TEXTOR. The DED generates externally applied magnetic perturbation fields which – so far unique – can rotate with a frequency of up to 10 kHz. The motivation for the construction of the DED was the possibility of an improved, less local energy and particle exhaust from the plasma and the more efficient shielding of the central plasma against impurities. In the meantime, many new problems have been identified, which can be investigated with the DED. This includes, in particular, the interaction of the magnetic perturbation field with the plasma and its in- fluence on the confinement and stability of the plasma. The selective excitement of tearing modes – current-driven instabilities that can also occur in ITER – has already led to the incorporation of TEXTOR into new joint experiments. With a view to the wall stabilization planned for ASDEX Upgrade aiming at higher plasma pressures for a possible steady-state operation, preliminary inves- tigations with the DED are planned at TEXTOR. The thematic link of the DED to stellarator phys- ics results from its three-dimensional magnetic field structure.
Another thematic priority at FZJ is plasma-wall interaction. The essential aim is to increase the life- time of the wall components in order to ensure sufficient availability for a future fusion reactor.
Jülich plays a leading role in the coordination of research activities at various experimental devices in Europe within the framework of the European Task Force on Plasma-Wall Interaction. FZJ has contributed its expertise in plasma-wall interaction to the application by the TEC partner FOM for a new linear experiment on the investigation of materials under steady-state and ITER-relevant plasma fluxes.
The most important results on the subtopics in the field of tokamak physics will be described in the following.
Improving the tokamak principle
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. In contrast to the stellarator, however, the toka- mak needs an internal plasma current for the generation of the confining magnetic field structure.
This involves the disadvantage for the tokamak that it has to generate and above all to maintain the plasma current. Up to the present, this has essentially been done inductively, which automatically limits the pulse length. The stellarator, which is in principle more suitable for continuous operation due to the lack of a plasma current, pays for this advantage with a very complicated coil arrange- ment. However, the inductively generated current in the tokamak can – at least partially – be re- placed by external or internal current drive sources.
This thematic area is much more explorative than the already far advanced development of the ITER reference scenario. Fundamental work on the magnetic confinement, on transport behaviour and stability is also combined here. The FZJ's major contribution consists in investigating the inter- action of the DED with the plasma. The DED was initially developed in order to additionally pro- vide the ergodic divertor concept with dynamic properties. The ergodic divertor is based on a per- turbation of the magnetic field at the plasma edge, so that a local heat load due to contact with the plasma wall can be spatially distributed over larger regions. If the perturbation field is additionally rotated, the loaded regions are spread even further. This basic effect was demonstrated using the
DED. Due to initial technical problems, however, it has not yet been possible to fully exploit the DED and its divertor properties.
Initial experiments with the DED on the stability of tearing modes, on angular momentum transport and on the density limit were unexpectedly successful. All three areas are of particular significance for ITER, since they are closely linked to the confinement and stability properties of the tokamak plasma and the resulting operational limits. Results that are to be highlighted are the positive effect of ergodic magnetic fields on plasma rotation, the controlled excitation of tearing modes and the dependence of the required perturbation amplitude on plasma rotation as well as the increase or decrease of the density limit – depending on whether the DED was operated dynamically or stati- cally. In cooperation with the Dutch TEC partner it was also shown that the tearing modes selec- tively generated by the DED can be stabilized again by local microwave heating. This procedure is to be used in ITER for stabilizing "neoclassical tearing modes".
Priorities for 2005 are the characterization of the ergodic divertor concept and possible conse- quences for an improved confinement, the comparison of the DED with the island divertor in the stellarator, the physics of the tearing modes and consequences for ITER as well as the connection between magnetic confinement and plasma rotation or momentum transport.
In the field of stability theory, the IPP carried out supporting investigations on the nonlinear cou- pling of external magnetic perturbation fields and plasma instabilities. Agreement with DED ex- periments was found for the coupling of modes with the same helicity. Future calculations at Re- search Centre Jülich will now concentrate on the dependence of the excitation of tearing modes on plasma parameters such as rotation or density.
Plasma-wall interaction
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 lining the plasma-facing components in ITER.
The investigations at TEXTOR are geared to these requirements: erosion and the associated reten- tion of the fuel (above all tritium), development of in-situ control and decomposition methods for limiting the long-term fuel inventory and qualification of the high-Z materials as alternative mate- rial. These issues are directly connected with the control of transient heat pulses occurring, for ex- ample, during ELMs or plasma disruptions, since these phenomena determine the power peaks hit- ting the wall.
Current research concentrates on the chemical erosion of graphite, on the resulting migration of carbon along plasma-wetted surfaces and in shaded regions and on the qualification of the in-situ decomposition methods for redeposited carbon. Experiments at JET have shown that long-range carbon transport takes place step by step, which requires a displacement of the contact areas be- tween plasma and wall. Reactive oxygen treatment is currently being tested in TEXTOR. Comple- menting the lining of ASDEX Upgrade with tungsten, unanswered questions are being dealt with at FZJ, which are associated with the use of tungsten components, for example, erosion at very high surface temperatures and the stability of molten tungsten layers under plasma load. Another impor- tant topic is the development of techniques permitting the time-resolved measurement of material deposition and tritium storage in ITER with the aid of laser desorption or laser ablation using a quartz microbalance.
For extrapolation to ITER this work is accompanied by the development of numerical codes. At FZJ, in particular, the ERO code incorporating an improved data base is used to predict erosion, redeposition and tritium retention in ITER.
Consolidating and extending the physical fundamentals for the design and operation of ITER
The reference operating scenario of ITER, on which the predictions for fusion power and power gain are based, is the H-mode. The H-mode is a robust confinement mode reproducibly achieved in all tokamaks with poloidal divertors such as JET and ASDEX Upgrade. For the preparation of ITER operation, a further improvement of the H-mode properties is being worked on. This includes the control of energy and particle exhaust, a further improvement of the confinement quality and an increase of the attainable plasma pressure. Important aspects are the avoidance of excessively high transient wall loads due to edge-localized modes (ELMs), as normally occur in the H-mode, and of pressure-limiting instabilities such as "neoclassical tearing modes".
According to the expertise of FZJ with the DED perturbation coil experiment, the Research Centre takes part in joint experiments at DIII-D (San Diego, USA) to suppress large ELMs by external magnetic perturbation fields. Initial encouraging results have already led to considerations concern- ing the application of this procedure at JET and later also at ITER. Further considerations now con- centrate on a better understanding of the stabilization mechanisms and on answering the question of whether this method is suited for application in ITER. For this purpose, FZJ participates in a feasi- bility study intended to clarify the installation of such perturbation coils in JET. Another option for avoiding large ELMs pursued by FZJ with work at JET is an H-mode with small ELMs (type III), which, however, also displays a lower confinement quality at lower target load. In support of this work the IPP is developing a method based on the control of ELMs by the injection of small frozen hydrogen pellets.
Radiation cooling of the plasma edge layer is a possibility of better distributing the heat diverted from the plasma to a few wall components. This procedure was further developed for JET. The es- sential element was the successful simultaneous control of radiated power and energy confinement time. The aim is to develop a robust control system for ITER.
Plasma diagnostics
The development of methods and techniques for measuring important plasma parameters and com- ponents influencing the plasma is an integral part of fusion research and thus an over-arching com- ponent of the programme topics tokamak physics, ITER and stellarator physics.
FZJ has a history concerning diagnostics development at TEXTOR going far back to the past. The same has been analogously applicable for some years to work for ASDEX Upgrade and JET. A major new development in the period under review was a quick-acting gas valve for JET to be put into operation in 2005. Ongoing developments are an imaging Bragg spectrometer, a VUV/XUV spectrometer for Wendelstein 7-X and a dispersion interferometer for interference-free electron density measurements in TEXTOR and later perhaps also in ITER. The Dutch TEC partners have successfully put an imaging electron cyclotron emission diagnostic device into operation, with which the electron temperature in the plasma can be measured in a two-dimensional cross-section.
Theoretical tokamak physics
The essential task of theoretical fusion and tokamak physics is the development of fundamental physical models to better understand and predict the processes in a fusion or tokamak plasma. Im-
portant topics are confinement and stability, the removal of energy and particles as well as plasma- wall interaction. The long-term goal is to make reliable predictions through reactor experiments.
Ongoing software code developments at FZJ include the ERO code for the description of plasma- wall interaction, the EIRENE code – a Monte Carlo simulation of the neutral particles in the plasma boundary layer – and the RITM code treating aspects of plasma transport. The EIRENE code was extended to cover molecular hydrogen physics and radiation transport. It was found that radiation transport is not negligible for the energy balance in the ITER divertor due to the high density and the large spatial extension.
In order to investigate magnetic perturbation fields as occur with the DED and their influence on radial plasma transport, the DALF3 turbulence code developed at the Max Planck Institute of Plasma Physics was extended together with this institute. Another joint project with the Max Planck Institute of Plasma Physics is the application of the coupled EMC3-EIRENE code to the DED. This code describes the plasma transport in a three-dimensional edge layer. The results thus obtained can later be transferred to the Wendelstein 7-X stellarator.
Stellarator
Research Centres Jülich and Karlsruhe as well as the Max Planck Institute of Plasma Physics in Garching jointly develop and construct the Wendelstein 7-X stellarator in Greifswald. The stellara- tor concept is regarded as an attractive candidate for a future fusion reactor due to its specific poten- tial for continuous operation. Wendelstein 7-X is a large stellarator, which has been optimized ac- cording to the quasi-symmetry principle. It consists of superconducting coils and is intended to pro- vide plasma discharges of 30 seconds duration at a heating power of 10 MW. The aim is to demon- strate the basic suitability of the chosen concept for magnetic confinement with long pulses.
According to its existing expertise, Research Centre Jülich has taken over comprehensive work packages for the construction of the Wendelstein7-X stellarator. This includes above all work for the design and manufacture 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 work described in the following was carried out in the year under review.
Construction of components for Wendelstein 7-X
For the superconducting bus system that connects the main coils with each other, an overall concept was developed which now comprises all steps of work and is intended to simplify assembly. In or- der to test the geometry of the bent conductors, a 1:1 model was created, which is now available at Jülich in a large hall for further work for conductor fabrication and assembly. The fabrication of the bus system will begin in spring 2005, so that delivery can take place in good time for Wendelstein 7-X.
The design of the joints needed for the electrical and hydraulic connection between the supercon- ductors was improved. Prototypes were then produced and successfully tested. The fabrication of all the 250 low-resistance joints will start at Jülich in 2005.
An analysis of mechanical stresses for various supporting elements was performed, which led to improved designs for some coil supports. Deformation analyses and stress calculations for the pla- nar coils helped to optimize the embedding process into the coil housing.
Welding tests at Jülich helped to obtain better predictions concerning material shrinkage and the associated accuracy to size during assembly of the coil supports. Some welding tests with proto-
types were carried out at Jülich. Advanced tests under more realistic conditions took also place at Greifswald.
Plasma-interactive materials
Dissimilar materials were subjected to a thorough analysis with respect to the mechanical stresses that occur during heating, in order to draw conclusions concerning their suitability as a material for optical windows at Wendelstein 7-X. The calculations show that for large windows up to a diameter of 13 cm sapphire is suitable for measurements in the visible to infrared region, even for the maxi- mum quasi-steady-state radiation exposure of 50 kW per square metre. Smaller windows made of ZnSe or ZnS with a diameter of 50 mm can be used for the far infrared region. Polycrystalline mag- nesium fluoride is suited up to diameters of 100 mm for thermal power densities of up to 20 kW per square metre only – higher power densities are possible with correspondingly shorter plasma pulses.
Calcium-fluoride and barium-fluoride windows cannot be recommended for applications with ther- mal load due to unacceptable stresses.
In order to investigate the properties of actively cooled components for the divertor targets of Wendelstein 7-X, tests with high heat loads were carried out in the JUDITH electron beam facility.
The aim of these experiments is to demonstrate that the heat removal capacity of the prototype tar- gets corresponds to the Wendelstein-7-X-specific loads of up to 10 MW per square metre.
Development of diagnostics
Detailed contracts were concluded for two diagnostic systems: for a VUV/XUV spectrometer sys- tem and a hydrogen diagnostic beam. The physical design of the HEXOS (high-efficiency XUV overview spectrometer) VUV spectrometer system and the procurement of the major components were completed. A subcontract for the development and production of the high-voltage supply sys- tem was awarded to the Budker Institute of Nuclear Physics (BINP) in Novosibirsk (Russia); the R&D activities for optimizing the grating structure and ion optics for higher current densities and lower beam divergences are currently being further pursued.
Additional diagnostics such as the imaging X-ray spectrometer, laser-induced fluorescence spec- troscopy, laser ablation and the thermal helium beam are currently further pursued within the framework of ongoing experiments at TEXTOR.
Further programme development
Fusion research at Research Centre Jülich is governed by two priority goals extending beyond the current HGF programme period:
a) Construction, commissioning (approx. in 2010) and scientific use of the Wendelstein 7-X stel- larator in Greifswald with Jülich participation.
b) Participation of Research Centre Jülich in the realization of the ITER project in Cadarache (France) and in the use of existing experimental devices for the optimization and preparation of the scientific use of ITER (from approx. 2013).
ANNUAL PROGRESS REPORT 2004/ASSOCIATION EURATOM − FZJ B. GENERAL PROGRAMME ON TEXTOR
B.1. PLASMA WALL INTERACTION
sçäâÉê=mÜáäáééë=EfmmF v.philipps@fz-juelich.de
Plasma-Wall Interaction (PWI) and its relation with the edge and general plasma behaviour is a ma- jor research topic at the TEXTOR tokamak. PWI processes determine the lifetime of wall compo- nents, the long term fuel (tritium) retention in the walls and affect the plasma edge such as the en- ergy exhaust by radiation. The experimental activities 2004 were strongly focussed on critical ITER PWI questions defined also by the work programme of the EU PWI Task Force. For ITER, the main open questions are the predictions of erosion and fuel retention caused by the use of graphite at the high flux area of the divertor, the development of control and removal methods to limit the long term fuel retention, the qualification of high-Z materials as alternative plasma facing materials and the control of transient power loads in ELMs and disruptions. Guided by these critical questions a main part of the activities was concentrated on the chemical erosion behaviour of graphite, the local and global carbon migration along plasma wetted surfaces and to shadowed areas (gaps) and the removal of redeposited carbon layers by oxygen treatment in TEXTOR. Another part of the work was focussed on experiments with high-Z tungsten (and Ta) wall components. Development of in situ techniques to measure material deposition and fuel retention in ITER was another main focus, followed up by further development of in situ laser desorption and quartz microbalance techniques (QMB). In addition the influence of erosion and material deposition on the optical properties of mir- rors has been analysed. This was accompanied by intensive modelling with the ERO impurity re- lease and transport code. Part of the work was performed on the JET tokamak within the framework of the Task Force E (exhaust and edge physics).
Hydrocarbon formation
The release of carbon has been intensively studied both in the JET divertor and in front of carbon and tungsten test limiters in TEXTOR. At JET a strong light emission of the C2Swan band and the CD Gerö band has been observed during strike point sweeps in the corner region of the inner MKII SRP divertor. The light emission could be correlated to the decomposition of probably soft hydro- carbon layers at the corner entrance. The spectroscopic observations are in line with measurements using the quartz microbalance located near the louver. Fig. 1 shows such molecular release around the strike point region in the JET divertor. By comparing CD and C2 photon emissions during known hydrocarbon injection into the outer divertor scrape off layer, a D/XB value for the Swan band could be determined which is in good agreement with the value measured formerly in the ex- periment PISCES-A.
Fig. 1: Emission spectrum from the inner divertor of JET at two different times during a slow strike point sweep over the horizontal and vertical divertor tiles. At 14.7 s the strike point touches the
corner region of the inner divertor, leading to strong CD and C2 emission. The inlet shows the temporal emission of the C2 light taken only from a wide and narrow band spectrometer.
Carbon transport and re-deposition
Carbon migration, re-deposition and material mixing were investigated by seeding 13C marked methane (13CH4 ) through graphite and tungsten spherical limiters. Shot-by-shot video recording shows a continuous growth of the deposit in the vicinity of the injection hole. Various post mortem analysis techniques – e.g. nuclear reaction analysis (NRA) and secondary ion mass spectrometry (SIMS) – showed that the fraction of re-deposited to injected carbon atoms are 4% and 0.3% for the graphite and tungsten surfaces, respectively, under identical seeding conditions. The maximum of the deposit is near the puffing hole for both limiters with a thickness being approx. a factor of two larger for the graphite than for the tungsten limiter. The large difference in the overall deposition ef- ficiency (more than one order of magnitude) is manly caused by the smaller deposition area on tungsten. The lower 13C deposition efficiency on tungsten surfaces is due to the different reflection coefficients (< 0.01 for carbon on carbon and ~ 0.4 for carbon on tungsten) together with an en- hancement of physical sputtering for carbon on a W surface due to the large mass difference. ERO and TriDyn code simulations are underway to separate and quantify these effects.
Carbon migration in TEXTOR was also investigated at the surface of the main toroidal belt limiter by a specially prepared and pre-characterized limiter tile having been exposed to the plasma during the 2003/2004 operation campaign for in total about 5000 seconds. Laser profilometry, SIMS and scanning electron microscopy (SEM) showed net erosion at regions close to the poloidal edge and significant deposition in the central limiter part. Strong deposition was also measured in the shad- owed areas of the limiter.
Carbon transport in the inner divertor of JET
The deposition of carbon at shadowed areas in the inner divertor of JET (louver area) was analysed further by the quartz microbalance diagnostic (QMB). In total, 806 exposures were done for a time of 6479 seconds in total under various divertor conditions. An average carbon deposition flux of 5.5×10-8 g/cm2s or 2.8×1015 C/cm2 was measured. Extrapolating this to the total operation time (26.4 hours) of the MKII GB SRP divertor (gas box divertor with septum replacement plate) yields a total amount of 35.4 g of carbon layers deposited on the louver region.
The deposition increases significantly with decreasing distance of the strike point position to the louver entrance. Elmy-H-mode discharges with the strike point on the horizontal target dominate the carbon layer formation on the QMB.
55000 56000 57000 58000 59000 60000 61000 62000 63000 0.0
5.0x1018 1.0x1019 1.5x1019 2.0x1019
Discharge number Carbon deposition (C/cm2 )
C
Fig. 2: Integrated carbon deposition on the QMB detector in the inner divertor of JET during the time period of March 2001 to January 2004. The detector was exposed to an overall plasma diver-
tor time of 1.8 hours while the total plasma divertor time during this period was 26.4 hours.
Carbon transport and fuel accumulation in gaps
In ITER the plasma facing elements will be castellated to insure the thermo-mechanical durability.
This increases the surface area and may lead to carbon deposition and tritium accumulation in the gaps in between cells. To investigate these processes a castellated test-limiter with a geometry as proposed for the vertical divertor target in ITER was exposed in erosion and deposition dominated zones in the SOL of TEXTOR. For the limiter exposed in the deposition zone, the averaged accu- mulated particle fluence was about 5×1019 D/cm2 and the limiter was investigated with various sur- face diagnostic techniques after exposure. Deposition was measured on the top plasma-facing sur- faces and in a narrow stripe-like zone in the gaps showing a maximum depth of about 370 nm and 200 nm, respectively. A conservative estimate gives a value of 30% of deuterium retained in depos- its in the gaps.
Fig. 3: View of the castellated limiter after plasma exposure in the deposition-dominated zone in the scrape-off layer of TEXTOR.
With the limiter exposed in an erosion dominated zone (particle flux about 8.6×1019 D/cm2), the plasma facing top surfaces in the erosion area were metallically shiny and they contained no meas- urable deposits. An intermixed metal-carbon layer was found in the plasma-nearest sides of the gaps with a thickness decreasing with increasing distance to the plasma, just like in the previous experi- ment. No deposits were found on those gap sides directly viewing the plasma.
Tritium retention and removal
One of the major issues in ITER and future nuclear fusion devices is the retention of tritium in re- deposited carbon layers. Methods to remove re-deposited carbon layers in situ that are compatible with ITER operation need to be developed urgently. Here, the removal of hydrocarbon layers by oxygen is one of the most promising methods that needs further development in various directions.
Also, ICRF discharges are considered as a promising candidates for wall conditioning and tritium removal – especially in superconducting fusion rectors.
A directly related pilot experiment has been performed in TEXTOR, addressing ICRF discharge initiation and a-C:H-film removal in oxygen/helium mixtures, employing about 15 pulses in total within the following regimes: PRF ≈ 110−120 kW, f = 32.5 MHz, τRF ≈ 1.0 s, BT = 1.8−2.4 T, O18/He4 ≈ 0.0−1.0, ptot = (1.5−3.3)×10-4 mbar. Thus, RF plasmas in a helium/oxygen mixture have successfully been produced and several observed effects should be mentioned:
0.4 0.6 0.8 1 1.2 1.4 1.6
0 0.2 0.4 0.6 0.8 1 1.2
0.4 0.6 0.8 1 1.2 1.4 1.6
0 0.1 0.2 0.3 0.4 0.5 0.6
Time [s]
Intensity [V]
Prfg [0.1 MW]
Vrfa [20 kV]
Rant [Ohm/m]
OI (LIMW3) Ha (LIMW1)
HeII (LIMW4)
TEXTOR #95654 (87%He + 13%O)
0.4 0.6 0.8 1 1.2 1.4 1.6
0 0.2 0.4 0.6 0.8 1 1.2
0.4 0.6 0.8 1 1.2 1.4 1.6
0 0.2 0.4 0.6 0.8 1 1.2
Time [s]
Intensity [V]
TEXTOR #95669 (0%He + 100%O) Prfg [0.1 MW]
Vrfa [20 kV]
Rant [Ohm/m]
OI (LIMW3)
HeII (LIMW4) Ha (LIMW1)
Fig. 4a: ICRF-DC in the gas mixture Fig. 4b: ICRF-DC in the gas mixture O2 : He = 0.13 : 0.87. O2 : He = 1.0 : 0.0.
• The flux of hydrogen atoms released from the wall was measured by spectroscopy to be propor- tional to the flux of oxygen atoms to the walls (figs. 7ab).
• An increased amount of hydrogen in the RF discharge with pure oxygen injection resulted in an improvement of the antenna-to-plasma coupling of up to ~ 50% (fig. 7b).
• The oxygen content in the RF discharges decreased very fast due to a consumption of oxygen by plasma dissociation and wall retention (in ~ 20−30 ms, fig. 7b). This effect limited the integral erosion of a-C:H-films and calls for further optimization.
• Normal tokamak operation was recovered fast and easily to standard operation after the oxygen RF discharges.
For comparison, glow discharge conditioning (G-DC) in the usual mode (current 8 A, pressure 7×10-3 mbar in continuous gas flow) was used in a mixture of oxygen and helium. The carbon mon- oxide partial pressure rises by a factor of 10 when G-DC is switched on, and the oxygen partial pressure drops by a factor of about 2.5. A-C:H films from pre-coated silicon probes with a thickness of 200 nm were completely removed after three hours of exposure to this glow discharge.
Hydrogen recycling
Fulcher-band spectroscopy has already been applied successfully in TEXTOR (see reports 2002 and 2003), Asdex-U and Tore Supra to determine the contribution of molecular hydrogen in the recy- cling hydrogen 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 di- vertor in L- and H-mode shots have been determined. The rovibrational population of the Fulcher bands of the deuterium molecules could be measured for the first five diagonal vibrational transi- tions by means of several high resolution spectrometers with lines of sight directed vertically and a spatial resolution from the outer vertical target plate to the centre of the MK IIB GB/SRP divertor.
Photon fluxes were converted into local molecular D2-fluxes using a collisional-radiative model showing that molecular sources contribute to about 70% to the recycling flux in the divertor. This was found to be in agreement with the neutral particle code NIMBUS.
Fig. 5: Ratio of protons from molecules to the total (proton) influx during JPN 61248 in the outer divertor. The shadowed area describes the variation introduced by the inaccuracies of S/XB- and D/XB-values. The rectangle is from EDGE2D/NIMBUS calculations.
High-Z limiter experiments
High-Z tungsten is foreseen at the upper divertor regions in ITER being a most promising candidate to replace CFC carbon also on the high heat flux lower divertor areas. Experiments in TEXTOR were concentrated on carbon re-deposition and re-erosion behaviour on W (already described), tungsten erosion at very high surface temperatures (> 2000 K) and on tests of a W-CVD coated bulk Cu limiter.
In the latter experiment, the CVD layer was created by chemical vapour deposition employing a re- duction process of WF6 gas on a Cu substrate at temperatures between 800 and 900 K with a depo- sition rate of about 0.2 mm/h resulting in a thickness of about 0.2 mm. The limiter was exposed to a high heat flux in the edge plasma of the TEXTOR tokamak.
Microscopic cracks occurred in the W layer due to the residual stress, but the layer showed good adhesion and no im- purity release could be identified except W and recycling O and C, confirming the excellent pureness of the layer. An esti- mated peak power density of 28 MW/m2 finally caused a destruction of the CVD- W layer initiated by cracking of the CVD-W layer, melting of the Cu sub- strate and ejection of liquid Cu through the cracks to the top, creating large cavi- ties under the CVD-W layer. In other ex- periments solid W plates were heated up by plasma impact until melting. The flux of tungsten atoms from the plate was measured spectroscopically in near UV region to reduce the influence of the thermal radiation con- tinuum coming from the hot surface. The intensity of a tungsten spectral line (W I, 284.8 nm) ver- sus the surface temperature is shown in fig. 6. No unusual increase of the W influx has been ob- served up to temperatures where normal thermal sublimation sets in. However, the activation en- ergy of the tungsten sublimation deduced from these data is about 6.7 eV, which is less than the heat of sublimation of tungsten. This may indicate the presence of some enhanced erosion.
2500 3000 3500 0
1 2 3 4 5
W : Eb=6.70 ±0.15 eV
WI(284.8nm) intensity / a.u.
T / K
Fig. 6: Tungsten W I line intensity versus the surface temperature of the tungsten plate.
Laser induced desorption in TEXTOR
In situ information on the hydrogen content of plasma facing surfaces is essential to estimate the long term tritium retention in future fusion devices. For this purpose a high power laser and spectroscopic detection of the released hydrogen has been implemented in TEXTOR (see also report 2003). The spectroscopic detection has been calibrated by targets pre-coated with an amorphous deuterium rich carbon layer (a- C:D) in a plasma deposition apparatus. Fig. 7a shows the measured hydrogen content of this layer and a carbon layer deposited during several TEXTOR discharges depending on the total plasma expo- sure. Due to the growth of the carbon layer, the hydrogen content increases with a rate of about 6×1015 D/cm2s. The deuterium content in the layer corresponds to a D/C ratio of ≈ 0.3.
In a laboratory experiment the desorption process was investigated on pre-characterised a-C:D lay- ers. Layers with thicknesses of 100 nm, 150 nm and 200 nm on a carbon substrate were exposed to
laser pulses of 80 kW/cm2 for about 300 µs and desorbed hydrogen and deuterium were observed in a quadrupole mass spectrometer (QMS) selecting the atomic masses 2 (H2 and D), 3 (HD) and 4 (D2). The laser radiation heats up the layer and a significant part of the bulk material underneath the spot. The D desorption scales nearly linearly with the layer thickness, whereas the desorbed H de- creases (fig. 7b). This finding might be explained by the diffusion time of H which is comparable with the laser pulse duration.
0 2 4 6 8 10 12 14
0.5 1.0 1.5 2.0
Desorption / 1018 (H+D)/cm2
number of tokamak discharges a-C:D film on graphit graphit
0 50 100 150 200 250 300
0.0 0.2 0.4 0.6 0.8 1.0
D H H+D
Desorbed particles / 1018 cm-2
a-C:D film thickness/ nm
Fig. 7a: Desorption signal from different spots versus the number of plasma discharges.
Fig. 7b: Desorbed D and H for different a-C:D layers on a carbon substrate.
Atomic and molecular data
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, databases (ADAS) etc. TEXTOR offers an excellent possibility to compare those theoretical data with experimental results. More refined S/XB values for B II have been obtained with code calculations and will be compared with experimental values derived from B(CH3)3 puffing.
Important tools (codes ATImpactParameterMethod [ATIPM] and ATCloseCoupling [ATCC]) for the calculation of atomic data were developed which extend the package “ATOM” developed ear- lier. These codes include the influence of heavy particle collisions and were used to interpret the line intensity profiles of helium atoms, which are conveniently used for the determination of elec- tron density and temperature profiles in the TEXTOR boundary layer. Fig. 8 illustrates the influence of the deuteron collisions. The left graph shows the density and temperature profiles from the model without heavy particle collisions leading to a non satisfactory fit of the line emission profiles and the right graph displays a much better fit obtained with a modified density profile (about 2 × ne- old).
In connection with the model for atomic helium, the time-dependent code formerly available in Jülich under the OpenVMS operating system environment has been ported to Unix and adapted to the evaluation of possible measurements with laser-induced fluorescence, both in the singlet and triplet systems of He I.
1400 1200
Fig. 8: Measured intensity profiles of three He I lines (singlet system: 667 nm, 728 nm;
triplet system: 706 nm) and the calculated fits (even curves) from ATIPM including deuteron collisions; particle penetration is from left to right.
Modelling
Carbon migration to the Quartz Micro Balance (QMB), which is installed at the entrance of the in- ner louver of the JET MkII-SRP divertor has been modelled with the Monte Carlo code ERO. The increasing deposition rate, when the strike point is moved downwards on the vertical divertor target, is in agreement with experimental findings. As in the experiments, the largest deposition on the QMB is seen with the strike point located on the horizontal divertor target. This configuration offers a direct line-of-sight for eroded particles to the inner louver region. A reasonable agreement of measured and modelled absolute deposition rates could be achieved taking into account an en- hanced re-erosion of deposited carbon layers. However, measured deposition rates show a large scatter indicating the influence of further parameters. Modelling suggests the formation of carbon layers on the horizontal target in configurations with the strike point located on the lower part of the vertical target (so-called DOC-LL configuration). The enhanced re-erosion of these layers can then lead to an enhanced deposition on the QMB when the strike point is for the first time located on the horizontal target. Such a dependence of the QMB deposition rate on the shot history is observed in JET.
Predictive modelling of target lifetime and tritium retention in ITER has been continued. A main uncertainty in the current modelling arises from the assumptions for the effective sticking (re- erosion) of hydrocarbons. First efforts have been started to analyse mixing effects by using a TriDyn based model.
Transport studies for chemically eroded hydrocarbons and physically sputtered carbon have been performed for the ITER divertor to study the effect of a dynamic change of the strike point position on the erosion/deposition pattern. Strike point sweeping leads to an erosion of formerly formed deposition layers and can therefore increase the target lifetime by a factor of about 1.5, being inde- pendent of the trivial geometrical effect.
A nozzle like geometry has been implemented into the ERO code for simulating dedicated hydro- carbon injection experiments at TEXTOR. The main aim is to analyse the influence of a material surface (surrounding the injection hole) on the hydrocarbon recycling and the resulting CD emis- sion. Work is underway to simulate D/XB values for different sizes of the nozzle surface and to compare them with experimental data.
-4 -2 0 2 4 6
0 200 400 600 800 1000 1200
X - Xlim / cm 728nm
706nm
667nm
-40 -2 0 2 4 6
200 400 600 800 1000
X - Xlim / cm
An ERO webpage has been installed as part of the homepage of the EFDA Task Force for Plasma- Wall Interaction (http://www.efda-taskforce-pwi.org/ero/). This page provides a new man- ual for the ERO code and also the ERO source code including the necessary atomic/molecular data and software for the evaluation of the ERO output files. A user friendly interface (“JERO”) based on Java has been developed to simplify the generation of parameter input files for ERO simulations.
A MatLab based software “MERO” has been developed for the visualisation and evaluation of ERO output.
ANNUAL PROGRESS REPORT 2004/ASSOCIATION EURATOM − FZJ B. GENERAL PROGRAMME ON TEXTOR
B.2. CONFINEMENT AND TRANSPORT
oçÖÉê=g~ëéÉêë=EfmmI=qb`LcljF r.jaspers@fz-juelich.de
In 2004 three existing topical groups, i.e. impurity transport, advanced tokamak scenarios and con- finement, were merged into the new group ‘Confinement and Transport’. This was done with the aim to promote internal discussion, to enhance the use of the available expertise, to focus the re- search on a few topics only, and to treat relevant aspects integrally. Although during this transition it took some time to form and settle the new structure, the following general aims and topics have been crystallized:
1. characterize the transport properties of the plasma in different regimes, 2. manipulate the transport properties by external means,
3. contribute to the ITPA joint experiments for situations where TEXTOR is unique or very well suited.
Ad 1) The transport and confinement in a plasma can vary strongly depending on which scenario is operated. For instance in the normal low confinement regime, the L-mode, the energy confinement is nearly independent from the electron density, whereas in the radiative improved mode, the RI- mode as discovered and explored at TEXTOR, the confinement is linearly dependent on density.
These differences are attributed to different turbulent transport mechanisms. The aim here is to measure and quantify the relevant turbulence and fluctuations in the different accessible regimes.
The various regimes explored at TEXTOR are: ohmic plasmas (no external heating), L-mode plas- mas with either neutral beam heating (NBI), ion or electron cyclotron heating (ECRH, ICRH) or a combination of these, the RI-mode in which a radiative mantle is generated by injection of a noble gas, leading to an improved confinement, the reversed shear scenario, in which the current profile is modified, such that the normally positive magnetic shear is reversed, and finally a regime where the edge of the plasma is ergodised by the external resonant magnetic field of the DED-system. For these investigations some unique diagnostic capabilities have been developed on TEXTOR, such as the ECE imaging system, reflectometry, multi-pulse Thomson scattering, and many more.
Ad 2) Once the transport mechanisms are known the ultimate goal will be to manipulate them, such that the confinement or transport can be increased – or decreased – on request. Methods to quench the turbulence, switch on or off transport barriers, ergodise the edge of the plasma etc. are being ex- plored with the TEXTOR versatile or specific tools such as the DED, the gyrotron system with the capability to drive current in the co- or counter direction, the co- and counter neutral beam injection etc.
Ad 3) An international tokamak physics activity (ITPA) has been established with the aim to pro- vide R&D support on crucial items and questions concerning burning plasmas. For the confinement and transport activity of this group, some collaborative joint experiments between various tokamaks
