Nuclear Fusion Programme – Progress Report 2009

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Nuclear Fusion Project: SC-FZJ 84(10)/4.1.2

Association EURATOM FZJ:  Annual Progress Report 2009

June 17th 2010

Member of the Helmholt

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edited by Ralph P. Schorn r.p.schorn@fz-juelich.de

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A. The Nuclear Fusion Programme of Forschungszentrum Jülich (executive summary) ... 5

B. Scientific and Technological Programme ... 15

B.1. Plasma-Wall Interaction ... 15

B.2. Tokamak Physics ... 43

B.3. Technology ... 67

B.4. Diagnostics and Heating ... 99

B.5. Contributions to ITER ... 133

B.6. Contributions to Wendelstein 7-X ... 138

B.7. Materials under High Heat Loads ... 139

B.8. Theory and Modelling ... 153

B.9. HPC-FF Operation ... 159

C. Specific Contributions of the Partners within the IEA Implementing Agreement ... 163

C.1. Japan ... 163

C.2. United States of America ... 172

D. Structure of the Fusion Programme and related Figures ... 177

E. List of scientific Publications, Talks and Posters ... 180

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Nuclear Fusion Programme – Progress Report 2009

A. Executive Summary

Ulrich Samm (KFS – Project Nuclear Fusion, u.samm@fz-juelich.de)

Introduction

Forschungszentrum Jülich (FZJ) as a EURATOM Association coordinates its fusion research activities within the Nuclear Fusion Project (KFS). The programme is based on several insti- tutes and is well embedded into the European fusion research structure, where FZJ is now fo- cussing on the two topics "plasma-wall interactions" and "ITER technology". The major part of the Jülich research activities is located within the Institute of Energy Research (IEF). The former Institute for Plasma Physics (now IEF-4 Plasma Physics) has by far 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 ITER development. IEF-2 (Microstructure and Properties of Materials) operates the high heat flux test facilities JUDITH 1 and 2 which are installed inside a hot cell and in a controlled area which is licensed to operate with toxic and radiating materials; this group represents the materials science expertise within the Jülich fusion programme.

The Central Technology Division (ZAT) provides engineering expertise and specialised work- shop capacities. The Jülich Supercomputing Centre (JSC) operates various types of supercomputer 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 expertises and to allow for the forming of a strong partnership as a consortium within the ITER construction phase. An updated TEC agreement with strong emphasis on the topic "plasma-wall interactions" and the joint use of dedicated facilities in Jülich, Rijnhuizen (NL) and Mol (B) is currently under preparation.

Co-operations beyond Europe are strongly supported by an IEA Implementing Agreement on

"Plasma-Wall Interaction in TEXTOR" together with Japan, USA and Canada, which mean- while also serves as a basis for the exchange of scientists to other devices than TEXTOR.

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Objectives and incorporation into the research area

Fusion research at Forschungszentrum Jülich is to a large extent scientifically organised along topical groups, i.e. plasma-wall interaction, tokamak physics, diagnostics, theory and modelling, and technology. These groups use a variety of different experimental facilities. Among these the most important machine is JET where scientists from Jülich are strongly involved, in particular in the scientific exploitation and also in the technical preparation of the new ITER- like wall project as well as in experiments addressing ELM-mitigation. Other facilities outside Jülich with participation from FZJ are DIII-D, PISCES-B, ASDEX-Upgrade, TS, LHD and MAST.

IEF-4 operates the TEXTOR tokamak as a local facility in Jülich (IP, max = 0.8 MA, BT, 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, ICRH 4 MW and ECRH 1 MW).

The Dynamic Ergodic Divertor (DED) on TEXTOR provides unique means for resonant magnetic perturbations: 16 helical in-vessel RMP coils; base modes: 12/4, 6/2, and 3/1, Imax = 15 kA, DC and rotating fields 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.

For Plasma-Wall Interaction (PWI) studies a powerful PWI test facility is available on TEX- TOR: two air-lock systems to expose movable and easily exchangeable larger scale wall components with gas feed, external heating and active cooling under ITER-relevant parallel heat and particle flux densities. The system is equipped with a comprehensive in-situ set of PWI diagnostics.

In addition the programme is supported by a variety of smaller laboratory devices: a tandem accelerator device for the quantitative determination of surface material compositions (NRA, RBS), dedicated laboratory devices for in-situ PWI simulation and analysis (TOF-SIMS) and various devices for the plasma assisted preparation of fusion relevant layers and coatings, and a

"mirror laboratory" for the characterisation and analysis of experiments with plasma facing mirrors in tokamaks.

The special expertise of IEF-4 in fusion technology is manifested by major engineering projects: concept development, design, construction and installation of the TEXTOR tokamak and various 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 tungsten 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. IEF-4 Plasma Physics has taken up substantial new projects for the development of ITER, based on special national funding. The task comprises R&D and design work for the CXRS diagnostic port plug system, the development of a new laser-based diagnostic system for Tritium retention, and the construction of a fast disruption mitigation valve.

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IEF-2 operates the high heat flux test facilities JUDITH 1 and JUDITH 2. These electron beam facilities are capable to deliver ITER- and DEMO-relevant quasi-stationary heat fluxes with loaded areas of up to 50 x 50 cm²and transient thermal load tests on a millisecond time scale with energy densities in the MJ/m² range to simulate Edge Localized Modes, plasma disruptions, and vertical displacement events. A unique feature of this test equipment is the operation inside a hot cell which allows testing of neutron irradiated and toxic materials (Be, T- containing samples).

ZAT is developing and manufacturing experimental devices and techniques for a wide range of scientific applications which are not available on the market. This central FZJ facility provides expertise in the fields of project engineering, joining and testing technology, and prototype manufacturing using special tools and techniques.

The Jülich Supercomputing Centre JSC operates a dual super computing system (both: general purpose and massive parallel architectures) and hosts the first dedicated European Supercom- puter for Fusion HPC-FF (100 Teraflop/s), which started operation in 2009 under an EFDA Implementing Agreement. HPC-FF is embedded into the European theory and modelling activities, such as the EU-ITM task force, and it also serves as a training platform for the Petaflop Computer for ITER, foreseen within the Broader Approach agreement between Europe and Japan.

The Helmholtz Association's fusion activities are based on the European fusion research programme. The following Helmholtz Centres are involved: Max Planck Institute of Plasma Phys- ics (IPP, Garching and Greifswald), Karlsruhe Institute of Technology (KIT, the former FZK), and Forschungszentrum Jülich (FZJ). Here, the research is organised along the topics: a) stellarator research, b) tokamak physics – ITER and beyond, c) fusion technology for ITER, d) fusion technology after ITER, e) plasma-wall interaction, and f) plasma theory. This report presents results having been achieved by FZJ in these topics.

Programme results

Stellarator research

Forschungszentrum Jülich is responsible for the design and fabrication of the superconducting bus-bar system and for some plasma diagnostic systems of Wendelstein 7-X.

Superconducting busbar system

The superconducting busbar system consists of a geometrically complex mesh of conductors being exposed to strong forces. It provides the electrical connections to the stellarator's magnetic field coils. After the set-up of the production line and the qualification of the fabrication and testing process, four out of five sets of busbars were manufactured, successfully tested, and delivered by end of 2009. Assembly of three modules is running in parallel in individual stages. The first module is complete and has been moved into the cryostat successfully. Busbar manufacturing is ongoing in a stable process. We intend to finish this task in June 2010.

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The design of the support structure is based on different adjustable sub-modules which are able to compensate fabrication tolerances in all directions and to facilitate the assembly on site. De- sign and stress calculations are taking into account these effects. The design of the supporting elements is finished. Supports for four out of five modules have been delivered to Greifswald.

The manufacturing for the remaining modules is ongoing. Modifications are done on request in case of clashes detected. Structural models of individual modules have been assembled to a global one validating the stress values initially computed.

Approximately 230 low-resistance joints are required for electrical and hydraulic interconnec- tions between superconductors at the coil terminals and between the five adjacent modules.

Based on a conceptual design for a pressure of 30 bars and a current of 18 kA a demountable joint for 200 bars and 20 kA has been redesigned. The manufacturing including inner clamping parts is ongoing at Forschungszentrum Jülich. Delivery is arranged according to the assembly schedule at Greifswald. A welding procedure was developed and tested for repeated mount- ability of the joints. Modified versions have been developed in order to facilitate assembly in bottleneck regions.

Diagnostics for Wendelstein 7-X

The test operation of the High Efficiency eXtreme ultraviolet Overview Spectrometer system (HEXOS) on TEXTOR has been continued with the aim to prepare and qualify the system for the later operation at the stellarator Wendelstein 7-X. Identification of spectral lines and impurity transport experiments have been completed, the development and testing of the software for the remote control system is in progress and the preparation of the specific mounting of HEXOS on Wendelstein 7-X has been started.

The hardware of four single-channel dispersion interferometer (DI) modules has been completed and in particular the safety and control system of the DI system was upgraded. A test operation on TEXTOR is under preparation in order to explore the technical applicability of the DI system for Wendelstein 7-X.

Tokamak physics

The tokamak physics programme of Forschungszentrum Jülich (FZJ) is a well focussed programme, which concentrates on three topics: resonant magnetic perturbations (RMP), plasma stability and turbulent transport.

Resonant Magnetic Perturbations

FZJ is highly involved in the development and understanding of ELM mitigation scenarios at JET and DIII-D. The ELM mitigation experiments using error field correction coils at JET have been continued with special emphasis on understanding and compensating the density pump-out, as well as further development of the understanding of the ELM mitigation mechanism with special focus on the scalability towards ITER. At DIII-D the work contributed by FZJ is focussed on the understanding of the edge particle and heat transport in a 3D topology by dedicated experiments under the leadership of FZJ and on accompanying 3D modelling

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activities using the code EMC3/EIRENE. These experiments are complemented by investigations at TEXTOR looking at the effect of RMP screening on the topology.

Plasma Stability

In 2009 a feedback control system for the control of tearing modes by Electron Cyclotron Resonance Heating (ECRH) has been implemented on TEXTOR. Control algorithms and hardware have been installed and taken into operation. The system combines a line-of-sight Electron Cyclotron Emission (ECE) diagnostic system as a feedback sensor employing a gyro- tron and a steerable ECRH launcher as actuators. The real-time detection and suppression of tearing modes has been successfully shown.

Massive gas injection experiments for disruption mitigation have been performed at JET under the leadership of FZJ. The gas was injected by a disruption mitigation valve (DMV) developed at Jülich, which was taken into operation at JET in late 2008. It was successfully shown that massive gas injection is able to reduce forces by halo currents, to dissipate most of the plasma energy by radiation and to suppress the formation of runaway electrons.

Turbulent Transport

The investigations on turbulent transport benefit from extensive diagnostic efforts to measure turbulence properties in the TEXTOR plasma. The work in 2009 concentrated on measuring the radial correlation and the long range correlation of the ambient turbulence and of the geo- desic acoustic mode (GAM) using correlation reflectometry, beam emission spectroscopy and Langmuir probes.

Fusion technology for and beyond ITER ITER diagnostics

Within 2009, substantial R&D and conceptual design work was performed on the ITER CXRS diagnostic system, based on special national funding. The work was organised within a consortium of associations, jointly led by Forschungszentrum Jülich (FZJ) and FOM/ITER-NL with participation of CCFE (UK) and HAS (Hungary). As a part of these activities, FZJ concentrated on the development and engineering analysis of prototypes for the CXRS port plug components (mirrors with mirror mounts, shutter, retractable tube, shielding cassette), as well as on experiments and modelling aiming at developing solutions to enhance the lifetime of the first mirror, which is the most vulnerable component of the CXRS system. Additionally, the development and testing of a prototype spectrometer (with ITER-NL) for the detection of the CXRS spectra is in progress.

In order to develop methods for the in-situ measurement of the stored amount of tritium on ITER, different laser-based spectroscopic approaches are under investigation at FZJ, including laser-induced desorption spectroscopy (LIDS), laser-induced ablation spectroscopy (LIAS) and laser-induced breakdown spectroscopy (LIDS). Within 2009, test experiments with LIDS were performed and laser systems for LIAS and LIBS were specified and procured. An additional

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smaller research activity addresses the development of a fast valve technique for the mitigation of disruptions in ITER by massive gas injection.

Materials for components in contact with fusion plasmas

A major objective of the R&D activities at FZJ is the characterization of plasma facing materials and components under quasi-stationary and transient thermal loads which are expected during the life cycle in future fusion devices such as ITER. These experiments are performed mainly in the high heat flux test facilities JUDITH 1 and 2 at FZJ using power densities of up to approx. 20 MW/m² (stationary loads) and up to the GW/m²-range (transient events), respectively. The research activities in the year 2009 have been mainly oriented towards the characterization of new beryllium grades with reduced oxygen content and the performance of mono- lithic tungsten grades (undoped grades and alloys) which are foreseen as wall armour for divertor applications. These materials have been exposed to intense thermal shock loads to simulate plasma disruptions and Edge Localized Modes (ELMs); main objective of these tests were the determination of threshold values for thermally induced surface modifications, crack initiation, and surface melting. In addition, actively cooled mock-ups with beryllium and tungsten armour were tested under cyclic thermal loads to quantify the thermal fatigue resistance of the interface between the plasma facing material and the heat sink.

In a joint initiative together with the Nuclear Research Institute NRI in Prague, First Wall qualification modules for ITER which have been provided by several ITER Domestic Agencies (EU, US, RF, KO, CN) according to the procurement plan were exposed to a pre-defined high heat flux test programme. In a first step steady state heating at 0.625 MW/m² was done in the BESTH facility at NRI for 30,000 thermal cycles. Afterwards the very same modules have been exposed to transient loading conditions (MARFE tests) at different power density levels ranging from 1.75 to 2.75 MW/m². Extensive diagnostics including thermo-couples, pyrome- ters, IR scanning and optical methods have been applied to detect thermally induced modifications and possible failure of the plasma facing beryllium tiles and the interface to the actively cooled heat sink.

To quantify the high heat flux performance of actively cooled divertor mock-ups with artificial defects, more than 100 mock-ups (CFC and tungsten monoblocks and tungsten flat-tile modules) had been produced by industry. After infrared inspection in the SATIR facility at CEA Cadarache and high heat flux testing in the electron beam facility FE200, all divertor mock-ups were investigated at FZJ by destructive metallographic examinations. The analyses were mainly focussed on the actual geometry of the artificial defect and on defect growth induced by thermally induced stresses during the high heat flux exposure in FE200. Besides a partial detachment of the plasma facing armour from the heat sink, erosion processes, crack formation and recrystallisation effects have been detected and quantified.

Plasma-wall interaction

Plasma-wall interaction research is the main focus of the R&D activity of the fusion programme at Forschungszentrum Jülich (FZJ). The research activities concentrate on the main and urgent critical ITER issues connected with the lifetime of plasma facing components and

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safety aspects. In more detail the activities are organised along the special working groups which have been established by the European Task-Force on PWI, such as material erosion and migration (addressing lifetime), transient events (lifetime), material mixing (lifetime and safety), fuel retention and removal (safety), metallic plasma facing components such as high-Z materials and liquids (mainly lifetime) and dust (safety). At present two of these groups are led by FZJ scientists (material migration and transients).

The FZJ PWI activities strongly concentrate on the JET tokamak within Task Force E, where Jülich has been intensively involved in 2009 with experiments on the previous carbon wall surroundings. FZJ is also strongly involved both in the technical preparation of the new ITER like wall (ILW) and in preparing the scientific exploration of the new wall starting early 2011, also by participating in the task force leadership. The design of the full tungsten bulk divertor row has been finished after a number of additional thermomechanical calculations and experimental tests and production is underway. In addition, the design of a number of diagnostic en- hancements such as the divertor endoscope, the main chamber camera systems and the new divertor Quartz Microbalance Systems has been finished and production started.

Material erosion and deposition is a key PWI issue. It has been investigated further by erosion measurements of graphite under low T plasma conditions (JET), by erosion of W limiters under TEXTOR edge conditions and by dedicated local impurity transport studies using local gas injections (hydrocarbons). Gaseous WF6 injection trough testlimiters has been used to study in detail the local W spectroscopy and the local W transport along surfaces. Additional dedicated material transport and deposition experiments have been preformed, in particular also in ITER like castellations, also in cooperation with DIII-D.

New wall conditioning methods for ITER under permanent magnetic fields have been followed up using ion cyclotron produced wall conditioning (ICWC) plasmas in TEXTOR and JET.

ICWC plasmas have been produced in a wide range of parameters in TEXTOR using the conventional ICRF antennas. Work in 2009 concentrated on hydrogen isotope exchange using H2

ICWC after TEXTOR walls have been saturated in standard GDC with deuterium. The isotope exchange efficiency has been optimized by varying gas mixtures, applying two RF frequencies and/or by overlaying a small vertical magnetic field to the toroidal field. ICWC plasmas in He/H2 mixtures are most efficient if oxygen is avoided. In addition, experimental studies on the removal of amorphous hydrogenated carbon (a-C:D) layers by glow discharge (GD) and electron cyclotron resonance (ECR) plasmas in reactive gases (H2, O2) have been performed in a simulation device (TOMAS).

The analysis of large Type I ELMs with very large energy losses of ΔWELM in the range 0.25- 1.3 MJ has been continued in JET. The ELMs provoke strong radiation losses, mostly confined to the inner divertor region. Large Type I ELMs with ΔWELM ≥ 0.72 MJ show enhanced radiation losses which are associated with the ablation of carbon layers in the inner divertor. Such large ELMs are often followed by a phase of Type III ELMs (so-called "compound" phase) with an increased radiation in the plasma core which lead to a plasma energy degradation.

R&D on W as PFM material concentrated also on the behaviour of bulk W under melting conditions in TEXTOR. The thermoelectric current from the molten W drives the molten W per- pendicular to the magnetic field line by j×B forces. Melt layer splashing has been observed

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with strong W plasma contamination leading to W accumulation and disruptions. Dedicated material analysis of molten W shows a dendritic like resolidification structure with a large po- rosity. Molten W droplets found to be redeposited on the non-molten surfaces show sizes between 2-6 microns.

Modelling

PWI R&D has been accompanied by intense modelling activities with the ERO code, concen- trating on material transport and deposition along plasma wetted surfaces in TEXTOR and inside gaps, modelling of erosion/deposition experiments with Be/C in Pisces, and chemical erosion studies in PILOT PSI at FOM. For Be it has been found that metastable states (MS) can have significant influence on light emission and on their transport.

Further ERO modelling of erosion and deposition in the divertor of ITER has been done for the initial ITER material mix of Be, W and C. In these simulations, new data on the surface temperature dependent tritium content in deposited carbon and beryllium layers along the divertor targets have been used. Parameter studies have been done by changing the plasma parameters and assumptions on enhanced erosion and sticking of hydrocarbon leading to variations in the retained T by about a factor of 2. Without any additional cleaning, 200 to 450 ITER discharges will lead to reach the safety limit of 700 g tritium.

ITER-like wall at JET

The ITER-like wall project (ILW) at JET addresses a number of specific and urgent PWI questions such as the long term tritium retention under ITER material conditions. It is also intended to investigate the compatibility of ITER operating scenarios with mixed metallic plasma-facing components (Be and W). The FZJ team plays a key role in supporting the ILW by leading the development and procurement of a full bulk W divertor row (ILT Project) and of improved diagnostics such as the divertor endoscope to measure W erosion (DES/KL11) or the quartz microbalance to detect material deposition (QMB).

The work in 2009 largely concentrated on completion of the bulk tungsten modules for the JET divertor. All manufacturing drawings are approved and fabrication of the bulk tungsten modules is well underway. A global thermal model of the bulk tungsten modules was used to calcu- late the temperature distribution on the surface and the W lamella supporting structures. The energy limit per W stack (4 in poloidal direction) is about 60 MJ/m². It is given mainly by the upper T limit of the Inconel wedge and the screws to fix the chain for the lamellas. From the engineering point of view, this implied finalising a very specific clamping arrangement. This was confirmed experimentally by exposure to a beam of ions and neutrals in the MARION test facility at Forschungszentrum Jülich.

Linear plasma devices: PSI-2 Jülich, JULE- PSI and MAGNUM-PSI

Plasma surface interaction in ITER and DEMO imposes new challenges with respect to particle and heat flux densities onto plasma facing components, both in steady-state and transients. This implies the use of toxic first wall materials and tritium (Tritium and Be in ITER), and the con- sideration of the impact of neutron irradiation onto first wall materials (DEMO)

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To characterize plasma surface interactions under these conditions, dedicated plasma surface interaction facilities are under development in the Trilateral Euregio Cluster (TEC: FOM in the Netherlands, FZJ in Germany and the Belgian Association). MAGNUM-PSI at FOM, a divertor simulator with superconducting magnetic field coils allowing for particle flux and power densities relevant for the ITER divertor, VISION-I, a compact plasma device inside the tritium lab at SCK-CEN Mol, capable of investigating tritium plasmas and moderately activated wall materials, JULE-PSI at FZJ, capable of exposing neutron activated and toxic wall materials to reactor relevant particle fluences and to ion energies including post-mortem analysis of neutron activated samples, and the high heat flux device JUDITH in the Hot Cells of FZJ for testing of activated and toxic samples.

The linear device PSI-2 has been transferred from Berlin to Jülich where it is presently rebuilt as a pilot experiment for JULE-PSI, which will later be operated in a hot cell. PSI-2 Jülich will be equipped with a new target exchange chamber which is currently being designed. The Hot Material Laboratory, where JULE-PSI will be installed in future, at present is under refurbish- ment to resume hot operation (also for the high heat flux experiment JUDITH-1 which will be moved to HML).

FZJ is responsible for the target exchange chamber and a number of specific diagnostics in Magnum PSI. The fabrication of the target exchange and analysis chamber has been completed. Laser techniques are foreseen to measure the hydrogen content and to characterise the material deposition via laser ablation (LIBS) and the QMB detectors which have been developed for JET will be used to measure deposition on remote areas also here. FZJ is also responsible for the implementation of the laser desorption, providing the laser and optical light guid- ing. LIBS laboratory measurements are underway at FZJ to qualify this method for application in Magnum PSI, PSI-2 and JULE-PSI.

Theory and modelling

Integrated edge plasma modelling

The theory and modelling activities at Jülich have been further refocused onto the outer domain of fusion plasmas, which are influenced by the presence of wall surfaces. This domain of theo- retical and computational activities reaches from "the barrier to the target".

Within the ongoing long term collaboration with the ITER edge modelling team the newest version of the 2D B2-EIRENE edge modelling code has been further developed. It was applied jointly with the ITER team, in particular for the divertor re-design according to the revised ITER 2008 design review constrains. The new divertor design (ITER reference F57) was significantly guided by these code calculations.

The large processor numbers and the nearly linear speedup of the EIRENE code on high performance computers such as the EFDA facility HPC-FF at Jülich enabled detailed and highly resolved studies of charge exchange spectra, resolved both in energy and angle, particularly at

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exposed surface segments in ITER, e.g. at diagnostic mirrors. This work is carried out under an ITER service contract.

European Transport Solver within the EFDA ITM Task Force

In the framework of activities of the EU Task Force for Integrated Tokamak Modelling a new integrated tokamak simulation code is envisaged: the European Transport Solver (ETS). For this package the fast module for neutral transport has been upgraded by including a description of D/T mixtures, and the exact (Monte Carlo based) module for the same issue (EIRENE) was made operational on the gateway computers. Furthermore, the RITM solver for transport equa- tions has been adopted for operation with arbitrary small time steps and installed as one of the solvers for the European Transport Solver (ETS).

3D fluid turbulence in the edge and SOL

The inclusion of self-consistent currents and electric fields is still one of the challenges in theo- retical and numerical studies of plasma transport. The plasma response due to currents arising in particular at resonant flux surfaces is of major importance for the suppression of intermittent transport and for the understanding of confinement improvement in the presence of resonant magnetic perturbations. These issues are addressed with the ATTEMP code.

In the framework of high performance computing (HPC) activities on the Jülich supercomputers, the ATTEMPT code was parallelized efficiently, allowing a speed up of two orders of magnitude in computation time for standard applications. The parallelised ATTEMPT code was embedded into the data structure of the EFDA Integrated Tokamak Modelling (ITM) project on the GATEWAY computer. The code structure was adapted to provide a turbulence module conformable with the CPO data transfer rules. It is now ready for routine applications in the KEPLER framework.

HPC-FF

The new EFDA fusion supercomputer HPC-FF – operated under an EFDA joint implementation agreement – was put into operation at the Jülich Supercomputing Centre JSC. On June 21st 2009 the combined system JUROPA/HPC-FF reached no. 10 in the world's Top 500 List and no. 2 in Europe with a performance of 308 Teraflop/s Peak, 274.8 Teraflop/s Linpack, while for HPC-FF itself the projected 101 Teraflop/s Peak and 87.3 Teraflop/s Linpack have been reached.

However, not until November 2009 the parallel file system Lustre – being a single point of failure – reached a sufficient degree of stability. Currently the utilization of both installations at FZJ's Supercomputing Centre – JUROPA and HPC-FF – is high. However, it took some research by JSC and ParTec to enable user codes to scale up to and above 1000 nodes. The so called cycle-swap mechanism, one of the key beneficial features for operating HPC-FF at JSC, used to exchange CPU cycles between HPC-FF and other JSC supercomputers, is also now in place: The fusion project "FSNGYRT" (multiscale gyrokinetic turbulence) benefitted from this option when applying for time on JUGENE (Jülich Blue Gene/P).

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Nuclear Fusion Programme – Progress Report 2009 B.1. Plasma-Wall Interaction

Volker Philipps (IEF-4 Plasma Physics, v.philipps@fz-juelich.de)

Introduction

Solutions of remaining critical issues of the next-step fusion device ITER, especially those related to plasma-wall interaction are needed in the current phase in order to support the engineering choices and to gain operational experiences to support a safe ITER operation. The activities of the Main Topic Group on Plasma-Wall Interaction (PWI) are strongly focused to these issues and remain organised in a programmatic oriented way to clarify open scientific questions and to develop also practical solutions to these questions. In this sense, The Main Topic Group is fully aligned with the goals of the European Task-Force on PWI (see http://www.efda-taskforce-pwi.org) for which ITER-relevant subjects are treated with high priority.

The strong contribution of the PWI Main Topic Group to the JET tokamak within the Task Force Exhaust is in line with the preparation work for ITER (see http://users.jet.efda.org/pages/e-task-force/index.html). An ITER like wall material combination has been prepared technically and from plasma scenario development viewpoint and will be tested in JET in the frame of the ITER-like wall project acting as a test bed for the currently foreseen material options in ITER. The continuous development of basic understanding of plasma-surface interaction and related near wall plasma processes is followed up in parallel.

Numerical modelling of material erosion, transport and deposition in TEXTOR, JET and other devices is an essential part of the work of the Main-Topic Group. The modelling is applied to provide predictions for key PWI questions for ITER operation.

The PWI work concentrates on three main open questions for PWI on ITER: (i) Safety issues with respect to tritium inventory, dust production and the lifetime of the plasma-facing components (PFC). This includes measurements and predictions of fuel retention, as well as techniques to clean up the plasma-facing components and control the dust inventory. (ii) Qualifica- tion of high-Z materials such as W for application in ITER and future devices with respect to a number of critical questions, such as erosion behaviour, melting, fuel retention, change of surface morphology etc. (iii) Radiation and heat loads under transient behaviour such as ELMs and disruptions and development of disruption mitigation techniques.

The Main Topic on Plasma-Wall Interaction deals mainly with the following fields:

(a) Carbon-based PFCs: Erosion, transport and deposition processes and fuel reten- tion in carbon deposits and bulk graphite. Emphasis is on erosion at ITER like divertor conditions & transport of carbon along surfaces. A special emphasis is on carbon transport to gaps and qualification of fuel removal techniques on top surfaces and remote areas such as gaps, also under the presence of magnetic fields.

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(b) High-Z PFCs: Research in this field concentrates on high temperature behaviour such as erosion, W melting, stability and droplet formation and on retention of hydrogen and helium in the bulk material.

(c) Material mixing: Research concentrates on mixing of W with C by codeposition and implantation.

(d) Transient heat loads: Radiation behaviour and heat load characterisation during ELMs and disruptions and development of disruption mitigation techniques.

(e) Plasma operation under detached conditions: Analysis of high density operation and divertor detachment in JET and under helical DED divertor conditions in TEXTOR.

(f) Qualification of atomic and molecular data: The study of plasma-wall interac- tion and the associated modelling requires a detailed diagnostic of the plasma boundary layer as well as good atomic and molecular database for the interpreta- tion.

(g) Development of modelling tools for erosion, deposition and fuel retention for ITER and other devices.

The Plasma-Wall Interaction group works on a number of facilities with the aim to take advan- tage of optimum conditions for the specific topic. The majority of the work is done at TEX- TOR and JET and additional contributions from AUG (ASDEX-Upgrade), DIII-D, Tore Supra, PISCES-B, Pilot-PSI, and others. The overall research programme is organised within the Tri- lateral Euregio Cluster, which encompasses the Belgian (ERM/KMS) and Dutch (FOM) partners in addition to the Institute at FZJ-Jülich. Partners from Japan, USA and Canada are closely linked to this research programme via the TEXTOR IEA agreement. TEXTOR has served as the central facility for these partners. Joint experiments are performed in the frame of different ITPA Divertor Scrape-Off Layer working groups.

1. Material erosion, migration and deposition

1.1. Comparison of methane and ethene injection through W and C limiters in TEXTOR

Experiments with ¹³C labelled ethene and methane injection through polished W and C spheri- cal limiters, installed at the same radial position in the lock system of TEXTOR and exposed to comparable plasma conditions (ohmic plasmas, Te = 48 eV and ne = 8 x 10¹⁸ m^-3 at the LCFS), were fully analysed and successfully modelled with the ERO code. The benchmark of the code was made by comparison of modelled and observed results of penetration depths of different (hydro)carbon species (CH, C2, C⁺ and C²⁺) determined via optical spectroscopy and ¹³C deposition efficiency on the substrates measured by NRA. Injections were made for each limiter in ten identical plasma discharges.

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47 48 49 50 51 46 45 44 43 42

LCFS

C Limiter

radius / cm

profiles 0 1 2

0 3 2 5 4 -5 -3 -4 -1 -2

CD A-X band 430.7+/-1.0nm intrinsic CD

# 103247 injection

hole

C H2 4

47 48 49 50 51 46 45 44 43 42

LCFS

W Limiter

radius / cm

toroidal direction / cm 0 1 2

0 3 2 5 4 -5 -3 -4 -1 -2

CD A-X band 430.7+/-1.0nm intrinsic CD

# 103217 injection

hole profiles

C H2 4

47 48 49 50 51 46 45 44 43 42

LCFS

radius / cm

0 1 2

0 3 2 5 4 -5 -3 -4 -1 -2

CD/CH A-X band 430.7+/-1.0nm

total CD+CH

# 103247 injection

C H2 4

C Limiter

47 48 49 50 51 46 45 44 43 42

LCFS

radius / cm

0 3 2 5 4 -5 -3 -4 -1 -2

CD/CH A-X band 430.7+/-1.0nm

total CD+CH

# 103217 injection

C H2 4

W Limiter

47 48 49 50 51 46 45 44 43 42

LCFS

radius / cm

0 1 2

0 3 2 5 4 -5 -3 -4 -1 -2

CD/CH A-X band 430.7+/-1.0nm extrinsic CH

# 103247 injection

C H2 4

C Limiter

47 48 49 50 51 46 45 44 43 42

LCFS

radius / cm

0 3 2 5 4 -5 -3 -4 -1 -2

CD/CH A-X band 430.7+/-1.0nm extrinsic CH

# 103217 injection

C H2 4

W Limiter

Fig. 1: CH/CD A-X light pattern in front of the actively pre-heated graphite and tungsten limiter. The reference phase, the ¹³C2H4 injection phase and the resulting difference signal are shown separately.

Figure 1 shows the observed radial distribution of the emission in the light of the CD/CH A-X band for ethene injection for W and C limiter. A reference phase prior to injection phase has been applied to determine the background light emission. The reference phase has been sub- tracted from the injection phase in order to obtain information of the break-up of the injected species. In the W limiter case also CH emission is observed without injection in the reference phase. This is due to transient layers, which appeared on the limiter surface caused by the deposition of the background carbon flux of about 3%. The overall difference of the intrinsic CH emission in the reference phase between C and W limiter is about 30% (lower for W).

The injection itself influences the flux balance at the limiter surface and modifies the properties of the transient layer. Substantial light emission of the C2 Swan band was observed during the injection of methane. Methane will not break-up in the almost undisturbed hot and ionising plasma and produce C2 molecules, therefore the higher hydrocarbon must be produced directly at the surface due to erosion. As the C2 emission is negligibly small in the reference phase and only appears substantial in the methane injection phase, it must be caused by the release of higher hydrocarbons in the transient a-C:H layer, which is formed during the injection.

The measured and modelled ¹³C deposition patterns and efficiencies for the graphite cases are depicted as an example in figure 2. The ERO code can well describe the substrate effect, which leads to a difference in the deposition efficiency of about a factor 2, and the difference between methane and ethene of 25% for C limiters and 50% for W limiters, respectively. The highest deposition efficiency was measured for heaviest hydrocarbon injection (ethane) in combination with the lightest substrate material (graphite). However, all deposition efficiencies are in the order of 1% and therefore the standard assumptions of low effective sticking for hydrocarbons (eff. sticking factor 0.15) and high re-erosion yields of 15% are needed in the modelling to re- produce the low deposition efficiency and the emission pattern.

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13C₂H₄injection through C limiter

13CH₄injection through C limiter

-0.5 0.0 0.5 1.0 1.5

Poloidal axis [cm]

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8750 1.000

-0.5 0.0 0.5 1.0 1.5

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8250 1.000

-0.5 0.0 0.5 1.0 1.5

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8750 1.000

-0.5 0.0 0.5 1.0 1.5

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8250 1.000

ERO ERO

EXP EXP

=2.1 %

=2.3 %

=1.7 %

=1.9 %

13C₂H₄injection through C limiter

13CH₄injection through C limiter

-0.5 0.0 0.5 1.0 1.5

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8750 1.000

-0.5 0.0 0.5 1.0 1.5

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8250 1.000

-0.5 0.0 0.5 1.0 1.5

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8750 1.000

-0.5 0.0 0.5 1.0 1.5

Toroidalaxis[cm]

0 0.1250 0.2500 0.3750 0.5000 0.6250 0.7500 0.8250 1.000

ERO ERO

EXP EXP

=2.1 %

=2.3 %

=1.7 %

=1.9 %

The appearance of the transient layers in the experiment and the need in the modelling to increase the re-erosion yields for redeposited layers in comparison with graphite substrate are consistent with each other.

Fig. 2: Comparison of experimental and modelled ¹³C deposition efficiencies for methane and ethene injection on polished graphite limiters under ohmic plasma conditions.

References:

Experiment: S. Brezinsek et al., Phys. Scripta T138 (2009) Modelling: R. Ding et al., PPCF (2010), accepted

1.2. Carbon erosion under detached divertor conditions in L- and H-mode plas- mas in JET

Power detachment is essential for the divertor target plates in ITER in order to sustain the high power load. Experiments in JET have been conducted in L- and H-mode to study the usually detached outer leg with respect to plasma parameters, incident particle and heat flux and in particular the chemical erosion of the graphite plate. Fig. 1 shows the temporal behaviour of CD, CII and D_during a density ramp in L-mode with transition from high recycling into detachment and, finally, approaching the density limit. The roll-over in the ion flux at the target is in-line with the increase of D which is due to recombination processes. The spatially resolved plasma parameters in front of the target have been deduced in-situ form the line ratios of high-n deuterium recombination lines and by Stark broadening (Te~2 eV, ne~1x10²⁰m^-3). The detachment starts at the separatrix and extends into the SOL. With increase of the degree of detachment a decrease of the carbon erosion has been observed. The reduction of the chemical erosion, marked by the CH emission and the hydrocarbon flux, cannot be explained solely by the reduction of the incident ion flux, but indicates also an energetic threshold for the chemical sputtering process at about Te~2 eV.

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OSP detachmnet extends in SOL PFR

Spatial distribution of H

Fig. 3: Time evolution of integral D_, CII and CD emission at the target plate in a density limit discharge with disruption at the end. Substantial decrease of CII can be observed as well as a dramatic decrease of CD, thus, chemical sputtering occurs with increase of the degree of detachment. Reduction in intrinsic CD emission is due to reduction of the source as extrinsic CH emission due to local methane

injection confirms.

Local methane injection was applied to ensure that the reduction in the light emission is corre- lated to reduction of the particle flux (Fig. 3). The inverse photon efficiency was determined under these detached conditions to 45+/-22. The net reduction of hydrocarbon flux from at- tached to detached conditions amounts a factor of 3.1.

Experiments in H-mode confirm principle observations made in the L-mode plasmas, e.g. that detachment starts at the strike-point and extends into the SOL during a density ramp as indi- cated in Fig. 4. Though a reduction in the physical and chemical sputtering can be observed, the absolute reduction is significantly lower in H-mode with ELMs. The chemical erosion is reduced in the phase between ELMs, where the plasma is detached and the electron temperature is low. However, substantial erosion takes place during the hot phases with ELMs burning to the target. The applied direct imaging divertor spectroscopy averages due to poor time reso- lution over the ELMs.

Fig. 4: Time evolution of the radial distribution of H during a density ramp discharge in H-mode. The strike-point was shifted to obtain better plasma parameters by Langmuir probes.

References:

L-mode: S. Brezinsek JNM 2009

H-mode: S. Brezinsek ITPA DSOL-2 report 2009

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0 1 2 3 4 5 6 7 Time [s]

1.050 1.072 1.094 1.116 1.138 1.160 1.182 1.204 1.226 1.248 1.270

S[m]

[1e20]

0.00 19.41 38.83 58.24 77.66 97.07

Valve T3

T3

Set of fibres for

Spatial information CH/CD

0 1 2 3 4 5 6 7

Time [s]

1.050 1.072 1.094 1.116 1.138 1.160 1.182 1.204 1.226 1.248 1.270

S[m]

[1e20]

0.00 19.41 38.83 58.24 77.66 97.07

Valve T3

T3

Set of fibres for

Spatial information CH/CD

1.3. Erosion of a-C:H layers induced by local ¹²CH4 injection in the ASDEX Up- grade divertor

Experiments with local injection of methane in the outer divertor leg of AUG have been performed in November 2009. The aim was (i) to explore the impact of a W surface, and therefore the hydrocarbon recycling, on the effective photon efficiencies of different (hydro)carbon tran- sitions (CH A-X band, C2 Swan and CII lines), (ii) to determine in-situ the layer thickness of deposited a-C:H on W, and (iii) to investigate the capability of strike-point sweeping for in-situ layer removal. The experiment was divided in different parts:

At first slow strike-point sweeping in L-mode has been utilised to determine the spatial extension of the intrinsic hydrocarbon emission zone. Deposition of CH layers was induced by local CH4 injection in the subsequent identical plasma discharges, where the plasma conditions (L- mode, ~0.5 MW ICRH) and magnetic configuration was comparable to previously performed

13CH4 tracer injection with normal Bt and Ip direction. A second slow strike-point sweep without local injection has been applied to determine the spatial extension of the extrinsic deposited carbon. As expected, the ¹²C deposition pattern determined in-situ in this experiment was similar to the pattern of ¹³C found in previous experiments. To disintegrate the a-C:H layer, H- mode plasmas have been applied with 7.5 MW additional power in order to utilise ELMs and the strike-point position for layer removal (Fig. 5).

Fig. 5: The spatial extension of the deposited carbon layer can be determined by the light emission of CH during the cleaning sweep (right). The deposition zone is comparable to the zone observed after

local injection of ¹³CH₄under similar plasma conditions in 2007.

A final strike-point sweep in L-mode has been used to determine the final extension and source strength of the remaining freshly build layer. Though a reduction of the carbon source has been detected, a complete removal of the topmost soft layer has not been observed. The layer has been observed in toroidal direction over several sectors and delamination of the a-C:H layer was observed at the edges of several tungsten divertor plates in subsequent discharges. The final analysis of the injected amount and the equivalent deposit is ongoing. Independently, a reference plate with deposited a-C:H layer on a tungsten was positioned in the divertor manipu- lator for the two H-mode discharges. Substantial erosion of the layer took place. The remaining layer thickness will be analysed in the laboratory.

References:

S. Brezinsek, ITPA DSOL-2 report 2009

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2. Fuel retention and removal

2.1. Material transport and fuel accumulation in castellated structures exposed under tokamak divertor conditions

In ITER, plasma-facing components (PFCs) will be castellated by splitting them into small-size blocks to maintain the thermo-mechanical stability. However, there are still concerns, especially on the performance of castellated structures under high heat loads and on retention of radioactive fuel and impurities in the gaps. The R&D program on ITER-like castellated structures is underway in TEXTOR addressing the physics and design issues of castellation.

Shaping of the castellation cells looks like a natural step to minimize the material transport into the gaps and therefore to mitigate the fuel inventory in the co-deposited layers. First investigations of an impact of shaping on carbon transport and fuel inventory in the gaps have shown only a moderate positive effect caused by shaped castellation cells on deposition and fuel inventory inside the gaps calling for future optimization of the cell shape [1].

Fig. 6: (a) A photo of the castellated sample exposed on DIII-D; (b) Layout of toroidal gaps. Locations of secondary ion mass spectrometry (SIMS) scans are shown with arrows; (c) Results of the SIMS pro-

filing of carbon. Measurements made along the arrows shown in fig. (b).

A new experiment was performed in the divertor of DIII-D, the National Fusion Facility in the USA, in the framework of TEXTOR - DIII-D collaboration. A castellated tungsten sample was installed onto Divertor Material Evaluation System [2] and exposed in the lower divertor for over 3 days of plasma operations under a variety of plasma conditions (Fig. 6a). The purpose of the exposure was to study carbon transport and hydrogen retention in gaps with varying en- trance geometry at shallow angles of magnetic field. After exposure several ion-beam and electron-beam surface analyses were carried out on castellation samples. The deposits with thickness of up to 250 nm were found in the poloidal gaps of castellated structures. Carbidization of the tungsten was detected at the interface between the deposited carbon film and tungsten substrate. The presence of carbide provides additional difficulties in cleaning of deposits inside the gaps. Strong anisotropy of deposition at toroidal gap surfaces was observed (Fig. 6b) similar to results from corresponding investigations on castellated samples exposed in the TEXTOR

0.0 0.5 1.0 1.5 2.0

0 100 200 300 400

500Intensity, a.u,

Distance along the gap, mm SIMS Data Right side

Left side

Carbon

R

Bt R

1^o-2^o

L

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boundary plasma [1]. According to the preliminary evaluation, carbon-tungsten mixing in the formed deposited layers both on toroidal and poloidal gaps is close to negligible. More detailed investigations are underway.

Studies of the castellated structures in DIII-D are performed in the frame of US-EURATOM collaboration program. Investigations of material transport and fuel accumulation in gaps are made in the frame of several tasks of the European Task Force on Plasma Wall Interactions:

WP09-PWI-01-02 and WP09-PWI-07-04 and are the subject of the IEA-ITPA joint experiments program, task DSOL 13.

References:

[1] A. Litnovsky et al., J. Nucl. Mat. 390–391 (2009) 556

[2] P. C. Wong, D. G. Whyte, R. J. Bastasz, J. Brooks, W. P. West and W. R. Wampler, J.

Nucl. Mat. 433 (1998) 258

2.2. Fuel retention and erosion under ITER-like mixed species plasma condi- tions in PISCES

Different CFC and fine-grain graphite grades have been exposed to plasmas containing (i) pure deuterium, (ii) deuterium and beryllium, (iii) deuterium, beryllium and helium, and (iv) deuterium, beryllium and argon. The experiments have been performed in the frame of a long-term secondment to the PISCES linear plasma device at the University of California San Diego.

Thermal desorption spectrometry (TDS) and nuclear reaction analysis (NRA) have been used to measure the amount and distribution of retained deuterium in the samples. For the case of pure deuterium plasma, parametric studies of deuterium retention in NB41 have been done with variations of the incident deuterium fluence ( = 110²⁵ – 510²⁶m^-2), ion energy (Ei = 20 – 120 eV) and sample surface temperature (Ts = 370 – 820 K). The results can be summa- rized as follows:

• In-bulk retention is similar in different CFCs and fine-grain graphites

• In-bulk retention is higher for lower exposure temperatures

• In-bulk retention scales as ^ with  depending on temperature:

 < ~0.5 for low Ts (surface diffusion along pores)

 =0 for Ts > ~800 K – saturation of retention for  < 310²⁵ D/m² (few sec of ITER pulse)

• In-bulk retention is higher for higher incident ion energies

• In-bulk retention is higher for lower fluxes

The addition of Ar and He to the D plasma has not resulted in an increase of the carbon erosion rate. Also, the admixture of Ar and He impurities did not produce a significant effect on the mitigation of chemical erosion of carbon in the Be containing plasma. The sputtering yield of Be by Ar is much lower than the sputtering yield of Be by D for the incident ion energies of

~10 eV and is of the same order of magnitude for energies of ~100 eV. Therefore, it cannot be expected that the addition of 10% Ar to the D plasma would significantly affect the particle balance in the surface interaction layer, thus preventing the formation of the protective Be car-

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bide layer. For He the situation appears to be similar, though it has somewhat higher sputtering yields than D.

References:

[1] A. Kreter et al., Fuel retention in carbon materials under ITER-relevant mixed species plasma conditions, Phys. Scr. T138 (2009) 014012

[2] A. Kreter et al., Mitigation of carbon erosion in beryllium seeded deuterium plasma under bombardment by argon and helium ions in PISCES-B, Proceedings of ICFRM 2009, Sapporo, submitted to J. Nucl. Mat.

2.3. Fuel retention measurement by laser induced desorption spectroscopy Laser induced desorption spectroscopy of retained fuel in wall components has been further developed in lab experiments and applied in TEXTOR. In the lab, the 2D pattern of the fuel retention of TEXTOR ALT limiter tiles has been determined by spot laser desorption. Laser desorption for W surfaces has been started in laboratory experiments on W samples loaded with hydrogen in a GDC discharge. Optimised laser parameters were determined to 0.5 ms at a power density of 1.5 GW/m².At a loading fluence of 2.2x10²³ D/m² at 525 K, the D retention in W has been determined to 0.02%.

3. Fuel removal and wall conditioning 3.1. Wall conditioning by ICWC

Wall conditioning for ITER under permanent magnetic field has been followed up using ion cyclotron produced wall conditioning (ICWC) plasmas in TEXTOR and JET. ICWC plasmas have been produced in a wide range of parameters in TEXTOR using the conventional ICRF antennas. Work in 2009 concentrated on hydrogen isotope exchange using H2 ICWC after TEXTOR walls have been saturated in standard GDC with deuterium. The isotope exchange efficiency has been optimized by varying gas mixtures, applying two RF frequencies and/or overlaying a small vertical magnetic field to the toroidal field. ICWC plasmas in He/H2 mixtures are the most efficient gas mixture (if oxygen is avoided). Fuel isotope exchange by ICWC is characterized by non saturated wall conditions under the comparable ICWC pulses in TEX- TOR leading to strong wall pumping in the initial part of ICWC application. The maximal initial hydrogen removal efficiency of ICWC is about 10¹⁹ H/sec and about 10 times below the removal efficiency obtained in H2-GDC. However, the ICWC removal efficiency increases with applied ICWC time indicating increasing wall saturation, while H2-GDC removal strongly decreases in time. Additional vertical magnetic fields increase the removal efficiency but only moderately in the range 10-30%. ICWC plasma temperatures and densities have been measured with Li beam technique and are in line with previous ICWC measurements using Lang- muir probes. In JET hydrogen isotope exchange with ICWC plasmas was tested in 2009 in a similar way by preloading the walls in standard H2-Glow Discharge Conditioning (GDC) prior to ICWC in (He+D2)-mixture or in pure D2. Low pressure and high power RF discharges were found to maximize H outgassing and minimize D retention. Gas balance after cryo-pumps re- generation shows that 10~20% of the short term retained H could be exchanged with D within