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 testprovid-ing 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 migradeposi-tion, tritium retendeposi-tion, 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
The magnetic topology of the DED forming a helical divertor has been proven. Based on the