Association EURATOM – FZJ Annual Progress Report 2012

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Foto: Harry Reimer

Association EURATOM – FZJ Annual Progress Report 2012

SC-FZJ 90(13)/4.1.2

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Nuclear Fusion Programme – Progress Report 2012 Contents

1. Introduction ... 3

• Objectives and incorporation into the research area

2. Plasma-Wall Interactions in Tokamaks ... 7

• Scientific Exploitation of the JET ITER-like wall

• Scientific Exploitation of TEXTOR

3. Materials and Components under High Heat Loads and Plasma Exposure ... 13

• Refurbishment of the Hot Materials Laboratory (HML)

• Technical developments for JULE-PSI at the pilot experiment PSI-2

• Characterization of materials and components under high heat loads o Characterization of carbon fiber composites for the ITER divertor o Nano-structured tungsten

o Different tungsten grades under transient heat loads at very high temperatures

o Qualification of ITER first wall component

4. Tokamak Physics ... 21

• Disruptions and Runaway Electrons

• 3D Field Physics for Transient Plasma Wall Interaction Control

5. Wendelstein 7-X ... 27

• Diagnostic alignment in 3D magnetic field topology

• Progress in Development of Key Diagnostic Components

• EMC3-EIRENE Modeling Guiding Diagnostic Developments

• Systems for Core Plasma Diagnosis Delivered to Wendelstein 7-X

• An Ion Cyclotron Heating System as collaborative TEC effort

6. Diagnostics for ITER ... 33

• Development of the ITER Core Charge Exchange Diagnostic system

• Development of methods for the measurement of Tritium retention within the first wall, material deposits and dust

7. Fusion Technology for ITER ... 38

• Overview

• Concept development for a plasma diagnostic system for ITER

8. Theory and Modelling ... 44

• Towards a new EU reference edge code B2.4-EIRENE

• Improvement of B2-EIRENE CPU performance

• Developments and application of EMC3-EIRENE code

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Forschungszentrum Jülich – Nuclear Fusion – Progress Report 2012 – Contents

• Localized sources and resulting spreading of impurities

• 3D drift fluid model for linear plasma devices (ATTEMPT code)

• Stochastic Differential Equations for Plasma Density Fluctuations

• Hamiltonian mapping techniques for test particle transport in stochastic magnetic fields

• TREE code simulation of plasma-wall interface

• Ongoing code projects and major plans for 2013

9. DEMO – The Route to A Power Plant ... 49

• Contributions to DEMO design studies

10. HPC-FF ... 52

• Short summary of status

11. Specific Contributions of the Partners within the IEA

Implementing Agreement ... 53

• Japan

• United States of America

12. Structure and Figures ... 63

• Forschungszentrum Jülich (FZJ)

• Trilateral Euregio Cluster (TEC)

• Fusion Programme of Forschungszentrum Jülich (FZJ)

• Relevant figures of the overall FZJ Fusion Programme

A. Publications ... 65

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Nuclear Fusion Programme – Progress Report 2012

1. Introduction

Introduction

The Helmholtz Association's (HGF) fusion activities are in line with the European fusion research programme. The following Helmholtz Centres are involved: Max Planck Institute of Plasma Physics (IPP, Garching and Greifswald), Karlsruhe Institute of Technology (KIT), and Forschungszentrum Jülich (FZJ). This report presents results having been achieved by FZJ in the year 2012.

Forschungszentrum Jülich as a EURATOM Association coordinates its fusion research activities within the Nuclear Fusion Project (KFS). The programme is based on several institutes and is well embedded into the European fusion research structure.

The major part of the Jülich research activities is located within the Institute of Energy and Climate Research (IEK). This is organized along a number of institute parts, among which fusion research is concentrated within IEK-4 Plasma Physics and IEK-2 Microstructure and Properties of Materials.

The IEK-4 Plasma Physics has the largest share of scientific staff in physics and technology for fusion, operates the TEXTOR tokamak, performs scientific work on JET and DIII-D, supports the Wendelstein 7-X construction and takes up significant projects related to the development of ITER. With the recent appointment of a new second director at IEK-4 (Prof. Linsmeier) it is intended to enhance the materials science expertise within the Jülich fusion programme and in particular in IEK-4: This will complement the activities in IEK-2, which operates the high heat flux test facilities JUDITH 1 and 2. They are installed inside a Hot Cell and in a controlled area which is licensed to operate with toxic and radiating materials.

The Central Technology Division (ZEA1) provides engineering expertise and specialised workshop capacities. The Jülich Supercomputing Centre (JSC) operates various types of super- computer systems, among which one device (HPC-FF) is dedicated exclusively to fusion research within EFDA.

The Association EURATOM-FZJ has very close contacts to the neighbouring EURATOM associations in Belgium and The Netherlands. In 1996 they together have founded the Trilat- eral Euregio Cluster (TEC) which provides a clustering of resources in order to perform a co- ordinated R&D programme, to operate or construct large facilities (TEXTOR, MAGNUM- PSI), to combine different kinds of expertise and to allow for the forming of a strong partner- ship as a consortium within the ITER construction phase. An updated TEC agreement with a

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Forschungszentrum Jülich – Nuclear Fusion Programme – Progress Report 2012 – Introduction

strong emphasis on the topic "plasma-wall interactions" and the joint use of dedicated facilities in Jülich, Rijnhuizen (NL) and Mol (B) has been signed in 2010.

Co-operations beyond Europe are supported by an IEA Implementing Agreement on "Plasma- Wall Interaction in TEXTOR" together with Japan, USA and Canada. In view of the fact that TEXTOR will be closed at the end of 2013 and that new opportunities are arising from new linear plasma test facilities, the parties of the Implementing Agreement have agreed upon a change of name and scope: “Implementing Agreement on the Development and Research on Plasma Wall Interaction Facilities for Fusion Reactors”. This change will involve a number of existing and planned linear plasma devices in Europe, Japan and USA. The start under the new name and scope is planned in the course of 2013.

Objectives and incorporation into the research area

Fusion research at FZJ is to a large extent scientifically organised along topical groups and projects. This report follows this scheme covering the topical groups and projects:

• Plasma-Wall Interactions in Tokamaks

• Materials and Components under High Heat Loads

• Tokamak Physics

• Wendelstein 7-X

• Diagnostics for ITER

• Fusion Technology for ITER

• Theory and Modelling

• DEMO - the route to a power plant.

TEXTOR and JET are the main facilities for the studies of Plasma-Wall Interactions in To- kamaks and Tokamak Physics. On JET scientists from Jülich are strongly involved, in partic- ular in the scientific exploitation of the new ITER-like wall as well as in experiments address- ing ELM-mitigation. FZJ operates the TEXTOR tokamak as a local facility in Jülich (I_P,max= 0.8 MA, B_T,max= 3.0 T, R = 1.75 m, a = 0.46 m, plasma volume 7 m³, circular cross section, toroidal graphite belt-limiter (pumped), 16 TF coils, pulse length 12 s; auxiliary heating power:

NBI co 2 MW, NBI counter 2 MW and ICRH 4 MW).

The Dynamic Ergodic Divertor (DED) on TEXTOR provides unique means for resonant magnetic perturbations: 16 helical in-vessel RMP coils with base modes of 12/4, 6/2, and 3/1, Imax

= 15 kA as well as DC and rotating fields of up to 10 kHz. Based on these means the programme participates in ELM-mitigation studies (joint experiments) and in the investigation of power exhaust in helical divertor structures in preparation of long pulse and steady-state operation in stellarators.

Two air-lock systems on TEXTOR provide a backbone for Plasma-Wall Interaction (PWI) studies. They allow exposing movable and easily exchangeable larger scale wall components.

The samples can be equipped with gas feed, external heating and active cooling under ITER-

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relevant parallel heat and particle flux densities. The system is equipped with a comprehensive in-situ set of PWI diagnostics.

As part of the re-orientation of the scientific program after shutdown of TEXTOR, the experimental means for research on Materials and Components under High Heat Loads will be enhanced significantly. Already existing facilities are the high heat flux test facilities JUDITH 1 and JUDITH 2 (ITER- and DEMO-relevant quasi-stationary heat fluxes on 50 x 50 cm²) based on electron beams and MARION, a 60 keV, 70 kW hydrogen/deuterium beam (15x10 cm² from 10 ms to 15 s duration). In addition we operate the linear plasma device PSI-2 which allows PWI research under steady state conditions.

In addition to these existing facilities, which already now provide unique features since the electron beam facilities are located inside a Hot Cell allowing the test of neutron irradiated and toxic materials (Beryllium and Tritium-containing samples), we are planning a number of new experimental devices. The most significant investment is a new linear steady state plasma device (JULE-PSI) inside a Hot Cell in the same building as JUDITH1/2. The combination of these test facilities in a common building together with post mortem surface and material diagnostics will form a new and powerful experimental infrastructure for this research field called the Hot Materials Lab (HML). The licensing procedure for the HML starts in the second half of 2013.

Moreover, fundamental aspects of erosion and hydrogen retention in ITER-related mixed materials will be studied in dedicated laboratory experiments outside the HML using accelerator- based techniques (e.g. RBS and NRA) and surface analysis techniques (e.g. XPS and TPD).

New experimental facilities are created at FZJ and dedicated preparation equipment is made available at the synchrotron HZB-BESSY II, Berlin. A strong new focus is established at FZJ, beginning in 2013, in the field of advanced materials for plasma-facing components for DEMO and a fusion reactor. The activities will be based both on material development and characterization, connecting the fusion activities with specialized FZJ institutes and external partners, e.g. in the fields of synchrotron tomography and micro-mechanical testing.

FZJ contributes to the scientific exploitation of Wendelstein 7-X based on the existing expertise in Jülich in the field of plasma-wall interactions. The general goal is to understand and qualify the island divertor including a coherent assessment of the long-term integrity of the first wall and divertor. A suite of diagnostics is considered. The conceptual designs will undergo review procedures at Wendelstein 7-X within 2013. Particular emphasis is put on diagnostics for edge transport studies for investigating particle and heat exhaust in the island divertor.

These edge diagnostics are accompanied by diagnostics for studying the core-edge interface transport and impurity levels. Those systems, i.e. HEXOS VUV spectrometer and X-ray spectrometer have already been delivered to and installed at Wendelstein 7-X.

Based on significant special funding by the German government, FZJ is working on R&D and prototype development related to Diagnostics for ITER and Fusion technology for ITER.

The special expertise of FZJ in Fusion technology is manifested by major engineering projects:

concept development, design, construction and installation of the TEXTOR tokamak including various major upgrades and recently the design, layout, manufacturing and assembly of the superconducting bus-bar system for Wendelstein 7-X, design and procurement for a bulk tung-

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sten plasma facing component for the new JET divertor and the design and procurement for the target station of the new experiment Magnum-PSI at FOM. The expertise in diagnostics is based on decades of experience with the scientific exploitation of TEXTOR.

The focus of new developments is on ITER core charge exchange diagnostics (CXRS), the development of laser-based Tritium Retention diagnostic techniques as well as disruption mitigation schemes. The main direction of this work comprises feasibility studies, conceptual design activities, engineering analysis as well as prototype development and testing of critical components, with substantial involvement of industrial contractors. Based on this special funding and supported by EU training programmes, enhanced university contacts and a systematic recruitment policy pursued by FZJ, a significant team of engineers and physicists has been formed, which is now taking up major work packages in the frame of grants by Fusion for Energy (F4E).

Theory and Modelling is an important part of the PWI programme. Apart from continued direct close collaboration with ITER IO, JET and EFDA associates, on integrated edge transport codes, such as B2-EIRENE (ITER), and EDG2D-EIRENE (JET), SOLED2D- EIRENE (CEA) the specific focus is mainly on studying erosion/deposition processes with a focus on T-retention by co-deposition under ITER-like material mix conditions and the particular geometrical boundary conditions (e.g. gaps). Plasmas with pronounced 3-d topology are addressed with the code package EMC3-EIRENE code (carried out jointly with IPP Greifswald) and 3D equilibrium including resistive current redistribution by HINT2 as well as self-consistent 4-field drift-fluid plasma response modelling by ATTEMPT.

In early 2011, a process has been launched under EFDA to establish a European DEMO working group with the aim to develop a conceptual design for a next step device relevant for the time after ITER (PPPT project). FZJ is involved in participation in this project within the main fields of expertise, in particular concerning the materials development, plasma-surface interaction studies, alternative target and first wall concepts and divertor modelling and system studies. Already since 2010, the German fusion laboratories IPP (Garching and Greifswald), KIT (Karlsruhe) and FZJ have entered into a close collaboration on DEMO related issues within the

“German DEMO working group”, which is organised within 13 topical groups (6 physics related, 7 technology related). In a series of six bi-annual meetings so far, involving about 50 scientists from the three labs, a wide range of aspects and research topics related to a DEMO fusion reactor have been broadly discussed and an outline of future elements for DEMO related research in Germany has been developed.

The following chapters provide more detailed information about results and developments from these topical groups and projects. Additional information is given about operation of the High Performance Computer for Fusion (HPC-FF) in Jülich and activities within the IEA Im- plementing Agreement on "Plasma-Wall Interaction in TEXTOR".

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Nuclear Fusion Programme – Progress Report 2012 2. Plasma-Wall Interactions in Tokamaks

Plasma-Wall Interactions (PWI) is one of the major critical issues on the way to a continuously working fusion power plant. The plasma-surface contact and the associated erosion processes leading to impurity production must be compatible with a sufficiently low core impurity concentration. The lifetime of wall components is determined by erosion and deposition processes.

The long-term retention of fuel is governed by PWI processes such as co-deposition and implantation, and defines for tritium the safety case of a plant. The present PWI tokamak activities in FZJ are largely focussed on the urgent PWI-questions as defined by the EFDA Task Force PWI and the JET Task Forces within the EU fusion programme, and the International Tokamak Physics Activity (ITPA) under the umbrella of the ITER organization.

In particular, the JET ITER-Like Wall provides the only integrated test bed to study PWI processes in the material mix in ITER: a full tungsten divertor and a beryllium main chamber. FZJ participated in JET work programme in form of the task force leadership, scientific coordina- tors of dedicated experiments and the scientific expertise in the area of PWI and power and particle exhaust. The key contributions listed below are related to work performed in this con- text.

Scientific Exploitation of the JET ITER-Like Wall

Plasma-wall interaction has changed significantly after the removal of all carbon based plasma facing components in JET (JET-C), such that the plasma with this ITER-like wall (JET-ILW) is now facing only components made of beryllium or tungsten.

In comparison it is clearly seen that with JET-ILW the retention of hydrogen and deuterium is reduced by at least one order of magnitude and the transport of eroded material from the main chamber into the divertor and in particular to the remote areas is also reduced, due to the absence of multiple-step transport based on chemical erosion observed in JET-C.

Also a strong reduction of divertor radiation has been found. This is a consequence of the absence of carbon radiation and the low radiation potential of beryllium and deuterium inside the divertor.

The particle recycling is increased causing a change in divertor characteristics with the consequence of an increase of the density limit by 30%.

The use of tungsten reduces the operational window of JET. For minimizing the W-source in the divertor operation at low plasma temperatures is required for achieving low impact energies of the impurity ions.

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Forschungszentrum Jülich – Nuclear Fusion – Progress Report 2012 – Plasma-Wall Interaction in Tokamaks

Impurity composition, beryllium sputtering and residual carbon content

The installation of the JET-ILW led to a reduction of the residual carbon content. The carbon influx at the edge plasma decreased by a factor of 20.

The campaign averaged carbon concentration reduced from 3.0% in JET-C to about 0.2% with JET- ILW. The average beryllium concentration with JET-ILW reaches about 1.6% which is lower than the corresponding carbon concentration with JET-C. This is caused by the lower primary erosion source at the Be-limiters due to the absence of chemical sputtering of beryllium (but with comparable physical sputtering of Be and C) as well as the absence of a primary Be-source in the divertor. The corresponding Zeff dropped from 1.9 to 1.2. The first wall re- mained intact over the first year of JET-ILW operation. Oxygen plays no role in the impurity content due to the good gettering properties of beryllium. Conditioning of the vessel was not required which is in contrast to JET-C where routinely glow discharge cleaning as well as Be evaporations were required. Limiter plasmas with low density and high temperature have shown significant beryllium sputtering at the limiters driven by self-sputtering. This behaviour has been reproduced by modelling with the ERO code. This is an important result for the prediction of Be-erosion in ITER. ERO modelling in the divertor is pending as it requires information from post-mortem analysis about the mixed material layers formed on top of the PFCs.

Long-term fuel retention

The main driver for the installation of the JET-ILW was the expected reduction in fuel retention. The expectations have been met: with the JET-ILW the long-term fuel retention has been reduced by more than one order of magnitude with respect to JET-C. This has been determined by global gas balances in various plasma and operational conditions.

Former predictions for ITER concerning the number of allowed discharges in full D-T operation (400s) before fuel removal techniques need to be applied have been to a large extend con- firmed by these measurements: extrapolation of tritium retention rates from JET results in high power H-mode leads to a maximum tritium retention rate of 3x10²⁰Ts^-1 in ITER and would allow about 1250 discharges before any cleaning is required.

The main mechanism for the retention are co-deposition with beryllium in the main chamber as and implantation in Be and W. In contrast to the long-term retention, short-term retention is slightly increased, however strong outgassing after the discharge compensates this effect. The latter provides a reduced wall inventory and allowed successful plasma break-down after disruptions, while this was hampered in JET-C by outgassing from carbon layers.

W-source, -transport and -control

Optical emission spectroscopy was used to determine the W-erosion.

Impinging Be-ions have been identified as the cause for W-sputtering at the outer divertor plate. This is the main tungsten impurity source for the plasma. The erosion yield with Be is

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about one order lower than in experiments with tungsten plasma-facing components and carbon as the main impurities (TEXTOR or ASDEX Upgrade).

The prompt deposition probability of eroded W in L-mode plasmas was determined to about 50%. High density and low temperature inside the divertor is required to minimize the steady- state W-source and avoid the risk of W-accumulation. In ELMy H-mode discharges, energetic ELM bursts (T_ion>50eV) drive impurities to the target plate. They contribute with about 80% to the total W-sputtering yield.

W-accumulation occurs in H-mode discharges in cases where (i) the tungsten source is too high or (ii) the transport to the core is unfavourable. An integrated scenario with strong W divertor- screening and absence of accumulation has been achieved with moderate gas fuelling (divertor cooling to reduce the source), high ELM frequency (min. 10 Hz to ensure impurity flushing) and with central heating (to expel W from the core). This scenario has been achieved in JET- ILW, but with a degradation of the pedestal leading to a loss of confinement by 20% with respect to unfuelled references in JET-C. Operation with moderate nitrogen seeding led to an unexpected recovery of the pedestal and increase of the confinement, but also to a W-peaking due to an increase of the W-source during the ELM bursts.

Disruption and disruption mitigation

The disruption behaviour with the JET-ILW has been analysed and compared with that under carbon wall conditions. It has been found that the absence of carbon as an intrinsic radiator has significant impact on the disruption processes. The radiation level during disruptions is much lower than under the old wall conditions. As a consequence the current quench rates are reduced and a significant amount of magnetic energy is being dissipated in form of heat flux to the main chamber wall. This has led to a significant increase of the peak temperatures at the Be-walls during disruptions and even melting at the plasma-facing components made of Be at the top of the vessel during uncontrolled vertical displacement events. This finding now calls for an efficient and reliable disruption mitigation system, in particular in view of the future ITER operation.

Divertor characterization and density limit

The reduction of the intrinsic carbon content by more than one order of magnitude in the plasma led, in the first place, to a reduction of the intrinsic divertor radiation by typically 30% and therefore to less cooling and a higher divertor plasma temperature with respect to JET-C at comparable plasma density. Beryllium cannot compensate the loss of radiation due to the lower radiation potential at typical divertor conditions. Impurity seeding (e.g. by nitrogen) is required to achieve a similar radiation fraction. In L-mode discharges with the JET-ILW deuterium contributes significantly to the total radiation in the divertor. In comparison to JET-C, the inner and outer divertor at JET-ILW are more symmetric with respect to detachment onset and the roll-over occurs at the same ion flux to the target. The absence of carbon radiation led to an

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increase of the density limit in L-mode by about 40% in dependence on the divertor configura- tion and allowed stable MARFE formation. Therefore, the operational window in semi or full detachment of the outer divertor leg is wider with respect to the upstream density and allows stable and controllable operation in this divertor regime as required in ITER. The density limit occurs at the same degree of divertor radiation; therefore it is likely that in the carbon device a radiation collapse determined the density limit. Initial studies in H-mode showed a similar beneficial increase of the density limit in H-mode (back transition to L-mode).

ELM control by application of error field correction coils

ELM control at JET by applying resonant magnetic perturbations (RMP) has been demonstrated with both, JET-C and JET-ILW. Depending on the collisionality, an increase in the ELM frequency of up to a factor of 20 has been found with moderate impact on the plasma confinement. Direct comparison of low and high collisionality discharges with n=2 RMP demonstrates the impact of this parameter on particle pump out. The amplitude of the magnetic signals lev- elled-off at a current strength of 2.4 kA in the external coils, and no further increase in the ELM frequency was seen above this saturation level.

Analysis of stability boundaries has shown that RMP application at JET induces edge safety factor resonant increase of the ELM frequency due to a shift of the stability limits due to peel- ing modes. This contrasts with the standard view on RMP ELM control, where mainly changes in pressure gradients, i.e. ballooning limits, are considered.

Scientific Exploitation of TEXTOR

Carbon erosion and associated ERO modelling

Tracer injection through test limiters exposed to the edge plasma of TEXTOR show in general very small local deposition efficiencies of the tracer species. ERO modelling can reproduce these deposition efficiencies only with the assumption of enhanced erosion of the deposited tracer species. Recent experiments with ¹³CH4 injection through roof-like polished graphite test limiters did reveal a clear dependence of the deposition efficiency on the injection rate. Where- as the ¹³C deposition efficiency is ~0.3% with an injection rate of about 10¹⁹ molecules/s, it increases by a factor of about 5 when the injection rate is lowered by a factor of 10. A new experiment with altered geometry of the test limiter has been performed. The methane is now injected through a cave-like structure – one aim of this experiment was to analyse the deposition from the injection at plasma-shadowed surfaces inside the cave. Post-mortem analysis did show that about 3% of injected ¹³C is deposited inside the cave and a surprisingly large amount (~9%) on the top surface of the test limiter. Thus, the overall deposition is at least 30 times larger than for the standard geometry. One possible explanation for the large deposition on top of the test limiter surface is a smaller injection rate per/area due to the cave structure. ERO simulations show in opposite to these experiments smaller deposition efficiencies with decreas-

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ing the injection rate, which can be explained with changing flux balance of depositing ¹²C and

13C particles. To reproduce the measured deposition efficiencies, ERO thus needs smaller enhancement factor for erosion of deposits for smaller injection rate. It is supposed that at low enough injection rate (and thus low enough depositing flux) no enhancement for erosion is necessary to be assumed. This will be studied in the future with further experiment and modelling.

Castellated structures for the ITER divertor

In ITER, plasma-facing components of the main chamber and the divertor will be castellated to improve the thermo-mechanical stability and to limit forces caused by induced currents.

Avoidance of melting of tungsten is of highest importance. On basis of previous results from TEXTOR, a new shape of the castellated cells was proposed and comparative modelling studies of conventional and shaped castellation were made. Thermal analysis was performed with the ANSYS code evidencing that the new shape allows the operation at 20 MW/m² of steady- state thermal load even for misaligned castellation. The complete suppression of the ion flux in the gaps of shaped castellation was predicted with particle-in-cell code SPICE2. Simulations with the Monte-Carlo 3D-GAPS code yielded at least an 11-fold decrease of beryllium content in the gaps of the shaped cells. The results for carbon deposition showed less pronounced effect of the shaping: a 7-fold decrease of carbon content was predicted. The difference is caused by the absence of chemical sputtering in the case of Be and reduced erosion of deposited material in the gap. Investigations have been started in order to qualify the new shape of the castellation and to validate modelling predictions. Experiments with identical sets of castellated W- structures were performed in TEXTOR and EAST, where the components were exposed inside the scrape-off-layer plasma, and in the outer strike point area of the DIII-D divertor. Each set comprised conventional and shaped cells allowing for direct comparison. Depending on exposure conditions, the carbon deposition was 1.9-2.3 times smaller in the gaps of shaped cells outlining advantages of the new shape of castellation.

Fuel removal of a-C:H layers by Ion Cyclotron Wall Conditioning (ICWC)

A quasi steady-state operation of ICWC in multi pulse mode (0.5 s pulse every 20 s) for a maximum of 1 hour was applied for the first time employing the toroidal field coils in CW mode at Bt=0.23T and coupling power in the high harmonics scheme. Values for the carbon erosion rate and erosion yields were obtained by exposing different laboratory samples and pre-exposed TEXTOR tiles on a newly developed probe and sample holder. The removal rates in O2 and D2

amount to:

• Total removal of layer of 420 nm within 300 s of D₂-Plasma at 350°C, corresponding to at least 84 nm/min or a removal rate of 5x10¹⁷ D/cm²min

• Removal rate for oxygen at 350°C was 38 nm/min or 2.4x10¹⁷ D/cm²min

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The hundred times higher flux of deuterium to the samples appears to overcompensate the ten times higher yield of oxygen. Thus, the removal of a-C:T by ICWC in deuterium is promising.

There is necessity to improve the ICR plasma production rates in oxygen to make it suitable for the application in a fusion reactor.

The issue of the inhomogeneity of ICRF plasma can be solved by extending the plasma production volume to the poloidal region of interest in the tokamak, e.g. the divertor, where most of co-deposition is expected. This has been achieved by applying vertical magnetic fields in TEXTOR. It has been further demonstrated that the recovery of plasma operation after O cleaning is possible in an all-carbon device by applying D²-ICWC, followed by He-ICWC.

Tokamak operation resumed without any further conditioning by, e.g., the glow discharge.

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Nuclear Fusion Progress Report 2012

3. Characterization of Materials and Components under High Heat Loads and Plasma Exposure

Refurbishment of the Hot Materials Laboratory (HML)

The Hot Materials Laboratory (HML) of Forschungszentrum Juelich was established in the 1960s in the frame of the research activities for the high temperature gas cooled reactor. After a refurbishment concerning the building and the supply units of the laboratories in order to allo- cate a laboratory for the investigation of radioactive materials for next step fusion devices during 2010-2011, in 2012 we completed the planning for the concept of a new PSI-lab inside the Hot Materials Laboratory. The main installation is a new linear plasma device (JULE-PSI) inside a Hot Cell, equipped with a target analysis and exchange chamber to load neutron activated material samples and to perform in-vacuo characterization of surface conditions and hydrogen content after exposure with laser based surface diagnostics (Laser induced ablation and desorption, laser induced break down spectroscopy). After a conceptual design review we plan to start the licensing procedure for the new PSI-lab in the second half of 2013. Additionally, outside of the Hot Cell but inside the controlled area of HML, a combined thermal desorption (TDS) and laser induced desorption (LID) device will be installed, the components of which are already delivered and fabricated.

Technical developments for JULE-PSI at the pilot experiment PSI-2

The optimization of the plasma source for JULE-PSI is currently being carried out on the pilot- experiment PSI-2 outside the controlled area of HML. For this linear device, also a prototype of the target exchange and analysis chamber has been designed and fabricated, which suc- cesfully was brought into operation in 2012.

The targets are exposed to the PSI-2 plasma by means of a manipulator system which allows a linear motion and rotation along and tilting across the magnetic field axis. This gives the highest flexibility for the orientation of the probe surface with respect to the diagnostic applied in the plasma and analysis chamber. Further features are an actively cooled target holder, the allo- cation of 5 thermocouples and 3 electrical connectors, and a gas supply. The whole manipulator is electrically insulated against the vessel ground to perform bias experiments with voltages up to 300 V.

Figure 1 shows a photo of the target holder and the head of the target manipulator inside of the PSI-2 target analysis and exchange chamber. The metal plate on top of the target holder serves as a mask to fix the targets and is designed for the individual size of the targets. The holder is

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Forschungszentrum Jülich – Nuclear Fusion – Progress Report 2012 – Materials

dovetailed to the manipulator head, a concept foreseen for JULE-PSI to allow fixing the target holder when loaded with activated targets by remote handling.

Fig. 1: Head of the linear target manipulator inside the new PSI-2 target analysis and exchange cham- ber

CHARACTERIZATION OF MATERIALS AND COMPONENTS UNDER HIGH HEAT LOADS

Characterization of carbon fiber composites for the ITER divertor

The work within the F4E grants GRT036 and GRT369 is performed in collaboration with NRG, Petten, The Netherlands. It is divided into two major tasks: 1) the thermo-mechanical, thermo-physical, microstructural and chemical investigation of two different 3-directional CFC (carbon fiber composites) grades, i.e., MEGGITT from Dunlop and an upgraded version of NB41 from SNECMA; 2) the thermal shock characterization of two tungsten grades, i.e., metal injection molded (MIM) tungsten and standard rolled tungsten.

In 2012, the three produced batches of the CFC grade MEGGITT were characterized and quali- fied by tensile testing and thermo-physical characterization in all three orthogonal directions, and chemical analyses showing no unexpected contamination. The tensile tests’ results show in Figure 2, beside the typical scattered data for heterogeneous materials, the linear relation between Young’s modulus and tensile strength, which is more pronounced for the x- and z- direction, while for the y-direction a stronger scattering is observed. The mechanical strength of the material is in all directions in accordance with the specifications set up by ITER.

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Furthermore, the measurements comprised the determination of thermal diffusivity and subse- quent calculation of the thermal conductivity by calculating the product of temperature dependent thermal diffusivity, specific heat, and density (Figure 3). Due to the large scatter in material density, the variation in thermal conductivity is quite strong and only for densities larger than about 1.8 g/cm3, the ITER requirements are met while below, depending on the direction, up to 30 - 70 % lower values were found.

Fig. 2: Tensile test results for the CFC MEGGITT from Dunlop in all three orthogonal directions:

dependence of Young’s modulus and tensile strength

The thermo-physical investigation consisted of the determination of the thermal expansion coefficient (CTE, Figure 4) via push-rod dilatometry. In accordance with earlier findings on NB31, the x-direction, consisting of strong pitch-fibers, shows the lowest expansion while the thermal expansion increases significantly in z-direction, made up by needled PAN-fibers. The increase of thermal expansion in x- and y-direction, when comparing the technical CTE from 30-1000 °C and from 400 – 1000 °C, shows that at lower temperatures the material exhibits the typical negative expansion coefficient.

0 50 100 150 200 250 300 350

0 50 100 150 200

UTS [MPa]

Young's modulus [GPa]

x-direction

Batch 2 Batch 3 Batch 4

0 5 10 15 20 25 30 35

0 5 10 15 20 25

UTS [MPa]

y-direction

0 5 10 15 20 25 30

0 1 2 3 4 5 6 7

UTS [MPa]

z-direction

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Fig. 3: Technical CTE from 30-1000 °C (left) and 400-1000 °C (right) as a function of density

Fig. 4: Thermal conductivity as function of density for all three orthogonal orientations

Nano-structured tungsten

The demands on plasma facing materials for ITER are rather challenging; however next-step fusion devices such as DEMO and future commercial fusion reactors will expose the materials to an even more demanding environment. An important remaining issue is the thorough investigation of different high potential material grades. Because of the currently unavoidable Edge Localized Modes (ELMs), thermal shocks are a significant factor in material exposure. Espe- cially since nano-structured tungsten can achieve an increased thermal shock resistance, it is an important category of enhanced materials.

This topic was further explored with different nano-structured tungsten grades that contained yttrium. In collaboration with the University of Science and Technology Beijing (USTB) small samples were manufactured that can undergo high heat flux testing at Forschungszentrum Jülich. Spark Plasma Sintering (SPS) was used as fabrication method to transform the fine grained powder mixture in solid material. This method should guarantee a sufficient amount of homogeneity and high density, while retaining a fine grain structure. The SPS-method, and more specific the temperature, pressure and sintering time used in the method, and the addition of yttrium, both affects the amount of grain growth that inevitably occurs.

0 0.5 1 1.5 2 2.5 3 3.5 4 4.5

1.72 1.74 1.76 1.78 1.80 1.82 1.84 1.86 1.88 CTE30-1000°C[K-1] x 10-6

Density [g/cm³] x-direction

y-direction z-direction

0 0.5 1 1.5 2 2.5 3 3.5 4 4.5

1.72 1.74 1.76 1.78 1.80 1.82 1.84 1.86 1.88 CTE400-1000°C[K-1] x 10-6

0 50 100 150 200 250 300 350

1.6 1.65 1.7 1.75 1.8 1.85 1.9

Thermal conductivity [W/m.K]

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Fig. 5: SEM picture (BSE) of SPS tungsten containing 0.25 wt% yttrium (left) and 0.75 wt% yttrium (right).

In the scope of this research, tungsten disks with a diameter of 20 mm and a height of 2.8 mm both with (up to 1 weight%) as without yttrium were produced. After characterization they were exposed to high heat flux testing. The characterization showed that all samples have a good density, higher than 95% of the theoretical value. Furthermore, an increase in yttrium content shows a clear correlation with micro hardness and a decrease in average grain size (see Fig. 5) up to 1.5 µm for the highest yttrium content.

High heat flux experiments were performed with the JUDITH 1 facility and a set-up with a Nd:YAG laser. The loading consisted out of 100 pulses of 1 ms pulses. This was done at room temperature with power densities between 0.37 GW/m² and 1.14 GW/m². Additional to these tests, certain samples were also loaded with 100 pulses of 1.14 GW/m² at an elevated base temperature of 400°C or with 1000 pulses of 0.37 GW/m² at room temperature.

Analysis of the resulting damage showed a performance increase at room temperature, with a damage threshold above 0.76 GW/m² for the tungsten grade with smallest grain-size. Such a high threshold signifies the possibilities for nano-structured tungsten. Also the crack statistics (see Fig. 6) and roughening parameter shows a positive influence on performance from yttrium doping.

Fig. 6: Analysis of the cracks (crack depth and crack distance) after e-beam loading in JUDITH 1 for tungsten grades with different yttrium content.

Nevertheless, a drawback is identified with the results at a base temperature of 400 °C. When tungsten is ductile at the base temperature, it does not crack at these pulse numbers and show only surface roughening. Since the materials with 0.75 wt% or more yttrium did crack at 400

°C, while a grade with 0.25 wt% yttrium was roughened, there are strong indications that there is an increase of the DBTT.

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Different tungsten grades under transient heat loads at very high temperatures

The thermal shock response of tungsten is strongly influenced by its temperature-dependent mechanical properties such as ductility, yield and tensile strength. Former studies focused on the thermal shock behaviour of tungsten at base temperatures of up to 700 °C. However, in some regions of the ITER divertor and especially in a fusion device like DEMO these temperatures will be much higher. In order to investigate the influence of these very high temperatures on the thermal shock response, different tungsten grades (pure W, W-Ta alloys) were heated to base temperatures of up to 1000 °C by a new boralectric heating element and exposed to thermal shock events with a pulse duration of 1 ms in the electron beam facility JUDITH 1. The development of the induced thermal shock damages, i.e. surface roughness due to plastic deformation and crack formation, were quantified at different power densities up to 1.5 GW/m² and up to 1000 pulses. Representative examples of the surface modification and damage formation after the thermal shock exposure in JUDITH 1 are shown in Fig. 7 for pure W with different grain orientations and in the recrystallised state.

Fig.7: SEM images of pure W after 100 (first row) and 1000 (second row) thermal shock events with an absorbed power density of 0.38 GW/m² and different microstructures: a+d) longitudinal, b+e)

transversal, c+f) recrystallised.

The obtained results give a good overview of the thermal shock response of different tungsten grades at a very high base temperature of 1000 °C and after 1000 pulses. These results in combination with the determined mechanical and thermal properties give information how they influence the damage behaviour of tungsten. WTa5 (5 wt% Ta) with longitudinal grain structure showed the best thermal shock performance in terms of plastic deformation and crack formation due to its high mechanical strength which also seems to compensate the significantly reduced ductility. A lower tantalum content of 1 wt% has also a beneficial influence on the mechanical strength, but this effect is not strong enough to withstand higher power densities or 1000 pulses. Severe damage formation was observed for the transversal grain orientation as well as the recrystallised samples due to the grain structure being prone to crack formation and its low mechanical strength. Especially for recrystallised materials the exfoliation of surface near grains due to thermal fatigue already after 1000 pulses is a severe problem for an application as PFM. Nevertheless, the results give information on the high temperature performance of tungsten under severe transient thermal loads and about the important material parameters that influence this behaviour.

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Qualification of ITER first wall component

Two small scale ITER FW mock-ups supplied by Fusion for Energy (F4E) are tested in electron beam facility JUDITH2 to determine the performance of the Upgraded Normal Heat Flux (U-NHF) design under thermal fatigue. The mock-ups are loaded cyclically (30 s on and 30 s off) under 2 MW/m² with ITER relevant coolant water conditions: in inlet temperature 70° C, pressure 2 MPa and coolant velocity 3 m/s. A picture of the MU is shown in Figure 8.

Fig. 8: Picture of a U-NHF-mock-up. A thin (c.a. 2 mm) copper layer is used to join each Be tile to heat sink made from CuCrZr which is attached to a stainless steel back plate.

The testing conditions and MU performance are under constant supervision. Single color and two color pyrometers are installed to measure the temperature of a predetermined area of the beryllium surface. An IR camera is installed to observe the temperature distribution of the full Be surface. Thermocouples (TC) are used to measure the bulk temperature of the Be tiles. The experimental data are recorded every half second via the well calibrated acquisition system.

For an absorbed power density of 2 ± 0.1 MW/m², which is corresponding to the testing re- quirement (deviation tolerance of 5%), the pyrometers indicate that the peak surface temperature of the beryllium surface is between 420 °C to 440 °C (see Fig. 9). An interlock system is set up between the IR camera and electron beam gun (EBG). Once a temperature above 600 °C is detected by the IR camera the EBG switches off automatically. The IR image in Fig. 9 shows an example when the switch-off event happened with the bright area at the Be2 corner indicating the overheating of the tile.

3-D FEM analyses have been applied to benchmark the experimental results. Due to the symmetric geometry, half of the major part of the MU is used as the simulation model. The utilized tool is Ansys. Figure 10 shows a temperature comparison between experimental and simulation results, indicating that the FEM calculated temperature has good agreement with the thermo- couple measured temperatures, while the surface temperature calculated by FEM method is around 60 °C lower than the experimental data. This discrepancy is supposed to be due to variations of the emissivity of the Be surface under electron beam exposure. Figure 9 shows the FEM calculated equivalent (von-Mises) stress distribution of the mock-up, which indicates that high stresses are located at the interfaces of Be&CuCrZr and Be&copper; the maximum equivalent stress (588 MPa) occurs at the joining interface between tile Be2 and the CuCrZr heat sink. The high stress concentration in this area is a clear indication that the bottom part of the beryllium tile might acts as a starting point for the failure. This prediction is in a good agreement with the experimental results (see the IR image in Fig.10).

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Fig. 9: Temperature comparison between FEM and experimental results.

Fig. 10: Equivalent (von-Mises) stress distribution calculated by FEM. IR image at the top-right corner indicates the failure of the MU in tile Be2.

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Nuclear Fusion Programme – Progress Report 2012 4. Tokamak Physics

Introduction

The activities of the Tokamak Physics Topical Group are strongly focused on areas where TEXTOR has unique characteristics and can make significant contributions for next generation fusion experiments, e.g. ITER, and improve physics understanding. The dynamic ergodic divertor (DED) is a unique tool to apply and study the effects of (rotating) resonant magnetic perturbations (RMP) with emphasis on the physics of field penetration, the applicability of RMPs to mitigate disruptions, the effect of RMPs on plasma rotation and edge turbulence, and the interaction of RMPs with plasma instabilities. Disruption studies in 2012 aimed at the understanding of runaway electron generation and loss processes, the characterisation of runaway electron beams, and the exploration of massive gas injection to suppress or mitigate the effects of runaway electrons. The good accessibility and a set of complementary turbulence diagnostics (electric probes, spectrometry, reflectometry) allowed detailed investigations of turbulence properties at the plasma edge.

Disruptions and Runaway Electrons

Magnetic Field Threshold for Runaway Electron Generation

Runaway electrons originate from the tail of the electron energy distribution. Normally, the energy gain due to the toroidal loop voltage is balanced by energy loss due to collisions. Be- cause the cross section for collisions decreases with increasing velocity, there exists a critical velocity above which the electrons gain more and more energy, they “run away”. In recent years it has been reported from a variety of different tokamaks that the occurrence of these runaway electrons after disruptions is more likely at higher values of the toroidal magnetic field.

This finding could be reproduced in a dedicated experiment on TEXTOR. The toroidal magnetic field has been varied in a series of shots with otherwise identical plasma parameters. Figure 1 (top) shows the runaway electron current measured after a forced radiative collapse by using a disruption mitigation valve to inject a large amount of argon into the discharge. The runaway current shows a clear threshold above which it increases with toroidal field, but below the threshold no runaway electrons are generated. The bottom of figure 1 shows the runaway current plotted as a function of the magnetic turbulence level during the current quench phase normalised to the toroidal magnetic field (note that the x-axis is scaled logarithmically). The obtained runaway shows a sharp drop to zero when a certain level of magnetic turbulence is exceeded. This measurements shows that the decrease in magnetic turbulence level with in-

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Forschungszentrum Jülich – Nuclear Fusion – Progress Report 2012 – Tokamak Physics

creasing toroidal field gradually suppresses losses of super-thermal electrons during the current quench and eventually (above threshold) allows for runaway electron generation.

Control of Runaway Electron Beams

High energetic runaway electrons can carry a substantial fraction of the pre-disruption plasma current. There is a huge potential to do serious damage (e.g. melting) when these runaway electrons hit the wall. Various means to suppress the generation of runaways are presently under investigation (see next chapter on massive gas injection). One possible option in case that complete suppression of runaways fails is to keep the runaway electron beam under control, prevent contact with limiters and wall, and apply suitable actions for a controlled de-confinement of the fast electrons. On TEXTOR some dedicated experiments were done in order to optimise runaway electron control ans understand spatial and temporal dynamics of the runaway electron beam.

Stable runaway phases of up to 70 ms have been obtained by stepwise optimisation of feed-forward control. These experiments have shown that below a minimum runaway current he beam develops a fast growing (ideal?) instability and gets quickly lost. Even during steady-state phases sudden (crash- like) loss events do occur.

Massive Gas Injection for Runaway Electron Suppression

The injection of huge amounts of noble gas using fast valves, so-called disruption mitigation valves, is one option to suppress runaway electron generation during the current quench of a disruption. The aim is to reach a plasma density which is high enough to increase the critical electric field for runaway generation above the actual field caused by the increase in loop voltage during the disruption. TEXTOR is equipped with two disruption mitigation valves of different size. The larger one is able to inject quantities of up to 11 barL of noble gases within a couple of ms. Standard measurements of the electron density (e.g. FIR interferometry) do not work reliably at these extremely high densities. TEXTOR is equipped with a dispersion interferometer based on a CO2-laser. Due to the principle of operation this type of interferometer is less prone to fringe jumps and offers the capability to measure plasma densities after massive

Fig. 1: Runaway electron current vs (top) toroidal magnetic field, and (bottom) magnetic turbulence level

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gas injection. First results show that the gas assimilation (i.e. the fraction of injected gas which is ionised) saturates with increasing number of injected atoms. This saturation seems to be consistent with the limited energy content of the plasma which does not allow to ionise all of the injected noble gas atoms. The determination of the effective density for runaway suppression, taking into account ionised and neutral atoms, suggests that the Connor-Hastie-Rosenbluth criterion has likely been fulfilled. Further experiments are planned for 2013.

3D Field Physics for Transient Plasma Wall Interaction Control

Introduction

In order to avoid damage to ITER’s divertor, the large heat fluxes associated with type-I ELMs must be either reduced or eliminated. One possible method of ELM mitigation or suppression consists of applying resonant magnetic perturbations (RMPs) to the plasma. In the vacuum approximation, these RMPs cause the formation of magnetic islands and stochastic regions in the plasma edge. This introduces a radial component to parallel transport and therefore enhances radial transport, which leads to density pump-out and consequently to a reduction in the edge pressure gradient below the threshold value required to trigger ELMs. However, this description is missing the effect of the plasma response to the RMPs. Current sheets can form on rational surfaces and produce a magnetic field that locally cancels out the external perturbation, thus screening the RMPs. One might expect that a plasma with screened RMPs would contain no magnetic islands or stochastic regions. Understanding the plasma response to RMPs is necessary for understanding the mechanism of ELM mitigation suppression using RMPs.

Mitigation of Type-I ELMs with n = 2 Fields on JET with ITER-like Wall

Mitigation of type-I ELMs has been observed with the application of an n = 2 field in H-mode plasmas on the JET tokamak with the ITER-like wall (ILW). Several new findings with the ILW have been identified and contrasted to the previous carbon wall (C-wall) results for comparable conditions. Previous results for high collisionality plasmas (�^*e,ped ~ 2.0) with the C-wall saw little or no influence of either n = 1 or n = 2 fields on the ELMs. However, recent observations with the ILW show large type-I ELMs with a frequency of ~ 45 Hz were replaced by high frequency (~ 200 Hz) small ELMs during the application of the n = 2 field. With the ILW, splitting of the outer strike point has been observed for the first time during the strong mitigation of the type-I ELMs. The maximal surface temperature (Tmax) on the outer divertor plate reached a stationary state and has only small variations of a few degrees due to the small mitigated ELMs. In moderate collisionality (�^*e,ped ~ 0.8) H-mode plasmas, similar to previous results with the C-wall, both an increase of the ELM frequency and density pump-out were observed during the application of the n = 2 field. There are two new observations compared to the C-wall results. Firstly, the effect of ELM mitigation with the n = 2 field was seen to saturate so that the ELM frequency did not further increase above a certain level of n = 2 magnetic

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perturbations. Secondly splitting of the outer strike point during the ELM crash has been seen, resulting in mitigation of the maximal ELM peak heat fluxes on the divertor region.

Measuring the Plasma Response to Applied RMPs on TEXTOR

On TEXTOR, the fast movable Mirnov probe can provide direct measurements of the magnetic field structure with applied RMPs from the Dynamic Ergodic Divertor (DED). Comparing the magnetic topology in the plasma edge with that in the equivalent vacuum case reveals the effect of the plasma response to the RMPs. Clear shifts in the phase of the magnetic field are found to occur on rational surfaces, indicating that screening currents are formed on these surfaces. For the first time, screening currents on multiple surfaces have been observed simulta- neously as shown in figure 1. Experiments of the dependence of the plasma screening current on the plasma rotation, heating power and plasma density with the FMMP have been performed on TEXTOR. The results show more screening effect observed with higher beam power, high density. Comparison of the experimental observations with modelling of non-linear MHD modelling has been done for understanding of the RMP filed penetration process. The modelling results agree well with experimental observations and explain the dynamic process of field penetration and the radial profiles of the surface currents measured.

Prediction of the Steady State Flow Induced by the NTV Torque on ITER

Intrinsic toroidal rotation due to NTV effect induced by three dimensional helical ripple in tokamaks has been investigated. The steady state flow due to NTV effect was self-consistently determined by searching the roots of the ambipolarity constraint. The magnitude of the steady state flow due to this effect is of the same order as the diamagnetic frequency. It can be in either co-current (electron root) or counter-current (ion root) direction in the low collisionality case, while it is only in counter-current direction (ion root) in the high collisionality case.

This method is applied for the modeling of intrinsic rotation induced by three dimensional effect in ITER. The trapped electron effect is found be really important in determining the intrinsic rotation in ITER. In a large range of plasma radius in ITER, there are three roots, which are similar to the stellarator case. The NTV torque drives plasma rotation in ITER towards one of the stable roots, depending on the initial condition. The magnitude of the steady

Fig. 2: Poloidal component oscillation of local field. The discontinuity at r~47 and 43 cm corresponds to the posi- tion of 4/1 and 5/1 surfaces.

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state flow in ITER is found to be around 1-3 times of the diamagnetic frequency in the modeling of both the baseline and steady state scenarios. Near the pedestal, the ion roots are close to 0, while the electron ones still scale like the diamagnetic frequency. The profile of the intrinsic flow also depends on the pedestal height.

Magnetic Topology Changes Induced by LHW and Its Profound Effects on ELMs in EAST

To date, in all existing RMP-ELM mitigation/suppression experiments the magnetic perturbations are induced by either in-vessel or external coil systems. In-vessel perturbation coils have been considered and designed for ELM control in ITER. However, in future fusion reactors, like DEMO, in-vessel perturbation coils may not be feasible. Thus, ELM control through actively changing magnetic topology by other mechanisms offers an attractive solution for next generation tokamaks beyond ITER.

Strong mitigation of Edge-Localized-Modes (ELMs) has been observed on EAST, when Lower Hybrid Wave (LHW) is applied to H-mode plasmas with Ion Cyclotron Resonant Heating (ICRH). This has been demonstrated to be due to the formation of Helical Current Filaments (HCFs) flowing along field lines in the Scrape-Off Layer induced by LHW. This leads to the splitting of the outer divertor strike points during LHW similar to previous observations with resonant magnetic perturbations. The change in the magnetic topology has been qualitatively modeled by considering HCFs in a field-line-tracing code.

Modelling of ELM with a Current Relaxation Model and Associated H-mode Experiments on TEXTOR

Further development of the relaxation based model for Edge Localised Modes has been done.

The distribution of the current density is investigated by the inclusion of a simple bootstrap current into the model and then varying the bootstrap parameters. A bifurcation of the relaxation widths is observed between cases where the relaxation includes the bootstrap current and those without. This could be the start of an explanation for the so called ‘compound’ Edge Localised Modes where multiple relaxations would take place in a very short period of time but only the first would include the bootstrap contribution. An analytical expression has been created for both bifurcation cases and compared with numerical calculations.

Investigate coupling between intrinsic rotation and turbulence-driven residual stress

Direct measurements of turbulent residual stress have been carried out in the TEXTOR tokamak by using a counter NBI torque to balance existing toroidal rotation in the plasma edge.

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Without NBI the parallel Mach number is not zero. The corresponding parallel-radial Reynolds stress is measured by a combined Langmuir and Mach probe array. Substantial residual forces are observed in a narrow region inside the limiter position, verifying the existence of the residual stress as a possible force for driving intrinsic plasma rotations. We have measured the residual force in a wide range of plasma parameters. The dependence of residual stress on the pressure gradient is consistent with the thermodynamic picture for the intrinsic flow generation.

The free energy of pressure gradient excites drift-wave turbulence, which further converts the thermal energy into kinetic energy of macroscopic flows. However, for high-density discharges this effect is strongly damped at a certain level.

Further studies suggest that such an impediment might be caused by a reduced Er×B shear rate which decreases the symmetry breaking mechanism for the generation of Reynolds stress.

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Nuclear Fusion Programme – Progress Report 2012 5. Wendelstein 7-X

Wendelstein 7-X is the largest stellarator in the world being built in Greifswald, Germany, by the Max-Planck-Institute for Plasma Physics, Greifswald. The mission of this helical advanced stellarator (HELIAS) type device is to show good energy and particle confinement with highly reduced net-currents in the plasma equilibrium obtained. This route promises an intrinsic stationary confinement of fusion relevant high temperature plasma. One key ingredient is the reliable and stable exhaust of energy and particles to maintain the first wall and divertor integrity and keep the plasma clean enough for sustained fusion power production. Wendelstein 7-X is the first step on this route, and the investigation of plasma-wall interactions and the qualification of the divertor scheme used is an important element of the research strategy.

The divertor concept used is the island divertor. It employs magnetic islands in the plasma edge, which represent the interface between the core plasma and the plasma facing components. The qualification of this island divertor concept is mandatory to explore the capability of a HELIAS type stellarator with island divertor as a candidate for a future fusion power plant.

To tackle this goal, a versatile diagnostic suite was defined and the realization is realized within a new founded working group structure on “Plasma Surface Interaction at Wendelstein 7-X”.

Diagnostic alignment in 3D magnetic field topology

The diagnostic systems envisaged as a systematic setup to contribute significantly to the divertor exploration at Wendelstein 7-X is depicted in figure 1. This analysis is result of a dedicated analysis of the magnetic field topology con- ducted in 2012. Local di- vertor spectroscopy em- ploying filter based camera systems allowing tomo- graphic reconstruction of emission lines as well as high resolution spectromet- ric analysis is envisaged as working horse for the characterization of the divertor regime. This includes spectroscopy on intrinsic species as carbon or hydrogen as well as on actively in- Fig. 1: Prioritized diagnostic systems and the systematics within

the magnetic topology of Wendelstein 7-X