Nuclear Fusion Project: SC-FZJ 82(09)/4.1.2
Association EURATOM FZJ: Annual Progress Report 2008
July 7th 2009
Member of the Helmho
edited by Ralph P. Schorn
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 ... 41
B.3. Technology ... 78
B.4. Diagnostics and Heating... 95
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 ... 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
Nuclear Fusion Programme – Progress Report 2008 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 Project Nuclear Fusion (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 largest part of the Jülich research activities is located within the Institute of Energy Research (IEF). The former Institute for Plasma Physics (now IEF-4) has by far the largest share of scientific staff in physics and technology for fusion, operates the tokamak TEXTOR, performs scientific work on JET and DIII-D, supports the Wendelstein 7-X construction and takes up significant pro- jects related to the ITER development. The IEF-2 (Microstructure and Properties of Material) 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 fusion programme. The Central Technology Division (ZAT) provides engineering expertise and specialised workshop capaci- ties. The Jülich Supercomputing Centre (JSC) operates various types of supercomputer sys- tems, among which one device (HPC-FF) is dedicated exclusively to fusion research.
The association EURATOM-FZJ has very close contacts to the neighbouring EURATOM as- sociations in Belgium and The Netherlands. In 1996 they have founded the Trilateral Euregio Cluster (TEC), which provides a clustering of resources in order to perform a coordinated R&D programme, to operate or construct large facilities (TEXTOR, MAGNUM-PSI), to com- bine different expertises and to allow 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” (with Japan, USA, Canada), which meanwhile also serves as a basis for the exchange of scientists to other devices than TEXTOR.
Objectives and incorporation into the research area
Fusion research at Forschungszentrum Jülich is largely organized along topical groups (plasma-wall interaction, tokamak physics, diagnostics, theory and modelling, and technology).
These groups use a variety of different facilities. Among these the most important machine is
JET where scientists from Jülich are strongly involved, in particular in the scientific exploita- tion and also in the technical preparation of the new ITER-like wall project as well as experi- ments on ELM-mitigation. Other facilities outside of Jülich with participation from FZJ are DIII-D, PISCES-B, AUG, TS, LHD and MAST.
IEF-4 operates the tokamak TEXTOR 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 mag- netic perturbations: 16 helical in-vessel RMP coils; base modes: 12/4, 6/2, 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 com- ponents 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 pro- jects: concept development, design, construction and installation of the TEXTOR tokamak and various upgrades, and most recently the design, layout, manufacturing and assembly of the superconducting bus-bar-system for Wendelstein 7-X, design and procurement for a bulk tung- sten plasma facing component for the new JET divertor and the design and procurement for the target station of the new experiment Magnum-PSI. Recently, IEF 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 develop- ment of a new laser-based diagnostic system for Tritium retention, and the construction of a fast disruption mitigation valve.
The institute IEF-2 operates the high heat flux test facilities JUDITH 1 and JUDITH 2. These electron beam facilities are capable for ITER- and DEMO-relevant quasi-stationary heat fluxes with loaded areas of up to 50 x 50 cm2 and transient thermal load tests in a millisecond time scale and energy densities in the MJ/m2 range to simulate Edge Localized Modes, plasma dis- ruptions, and vertical displacement events. A unique feature of this test equipment is the opera- tion 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 which are not available on the market for a wide range of scientific applications. This central facility of FZJ provides engineering expertise in the fields of project engineering, joining and testing technol- ogy, and prototype manufacturing using special tools and techniques.
The Jülich Supercomputing Centre operates a dual super computing system (both: general pur- pose and massive parallel architectures) and hosts the first dedicated European Supercomputer for Fusion (100 Teraflop/s), which will start operation in 2009 under an EFDA Implementing Agreement. This HPC facility 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 Com- puter 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 pro- gramme. The following Helmholtz Centres are involved: Max Planck Institute of Plasma Phys- ics (IPP, Garching and Greifswald), Forschungszentrum Karlsruhe (FZK), and Forschungszen- trum Jülich (FZJ). The research is organized along the topics: a) stellarator research, b) toka- mak 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 mag- netic field coils. After the set-up of the production line and the qualification of the fabrication and testing process, three out of five sets of busbars were manufactured, successfully tested, and delivered by end of 2008. The final assembly of module 5 and the preassembly of module 1 have been completed in 2008. Busbar manufacturing is ongoing in a stable process.
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 modules 5 and 1 have been delivered to Greifswald. The manufacturing for the remaining modules is ongoing.
Approximately 230 low-resistance joints are required for electrical and hydraulic interconnec- tions between superconductors at the coil terminals and between 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. After design review three joints have been manufac-
tured and tested under pressure. 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.
Diagnostics for Wendelstein 7-X
The new High Efficiency eXtreme ultraviolet Overview Spectrometer system (HEXOS) for the stellarator Wendelstein 7-X is currently being operated on TEXTOR for a testing and qualifica- tion period. An absolute intensity calibration, the identification of spectral lines and dedicated impurity transport experiments have been performed successfully, and the development of a full remote control system is in progress.
A high energetic hydrogen beam (RUDIX) is jointly developed by IPP Greifswald, For- schungszentrum Jülich and BINP Novosibirsk for plasma diagnostic applications at Wendel- stein 7-X. Ion temperatures profiles are measured by neutral particles analysis (NPA) and charge exchange recombination spectroscopy (CXRS) that require a beam of 60 keV particle energy and 5 A equivalent neutral current. The BINP Novosibirsk has started the fabrication of the major components, the high voltage power supply, the neutralizer chamber, the calorimeter with beam dump and the radio frequency ion source. For RUDIX the concept of local control and its integration into the Wendelstein 7-X main control system has been compiled.
The performance of a spectroscopic analysis system for beam parameter characterisation fore- seen at RUDIX is currently being tested at the TEXTOR diagnostic beam.
The in-vessel optical components for a 12-channel dispersion interferometer (DI) have been installed at TEXTOR. The long term transmission behaviour of mirrors and retro-reflectors under plasma operation is currently being investigated.
Tokamak physics
The scope of this topic includes the physics being related to Resonant Magnetic Perturbations (RMP), plasma stability and edge transport.
In 2008 the activities on ELM mitigation and suppression at JET as well as the collaborative experiments with DIII-D have been continued. The JET activities concentrated on the exten- sion of the parameter space for ELM mitigation with special focus on particle fuelling. The collaboration with DIII-D is focused on understanding the transport processes during RMP, comprising a joint experiment in DIII-D and TEXTOR, which is embedded into ITPA. These RMP activities are complemented by studies on turbulence properties (GAM and intermittent SOL transport), field penetration (including studies on tearing mode excitation) and the loss of fast ions by Collective Thomson scattering (CTS).
Work on disruption mitigation by massive gas injection (MGI) was continued with the focus on the exploitation of different gas species and the suppression of high-energy resp. runaway electrons. The latter was also successfully addressed with RMP. At JET the new Disruption
Mitigation Valve (DMV) was taken into operation in December 2008 and first experiments were performed under the leadership of FZJ. Work on both machines, JET and TEXTOR, is part of ITPA coordinated efforts. Further development of the DMV towards ITER-relevant injection rates was continued in 2008.
Experiments on the active control of tearing modes by ECRH have been performed in 2008 in preparation of a feedback system, which will be commissioned and operated in the first half of 2009.
High density regimes, especially detached regimes achieved with the dynamic ergodic divertor (DED) have been studied using new spectroscopic diagnostics in 2008. Because of the 3D magnetic topology, this work is highly related to stellarator divertor physics. Different con- finement regimes with the DED have been studied, namely a density pump-out and an im- proved density confinement regime.
Suppression of high energetic electrons in plasma disruptions
Unstable conditions in tokamak plasmas can cause a fast uncontrolled current shutdown, a so- called disruption. During such a disruption, the high toroidal electric field can lead to the gen- eration of high-energetic electrons, the runaway electrons. Especially in ITER massive run- away generation is expected, with currents of up to 10 MA carried by these electrons. When lost to plasma facing components, these electrons can cause extreme heat loads, which signifi- cantly reduce the life-time of wall elements. Thus, the suppression of runaway electrons is an essential task to be solved for ITER.
Presently, massive gas injection is considered as a tool for runaway suppression. However, the amount of gas and the timescale for the gas injection is extremely challenging. An alternative method might be the suppression of runaway generation by increasing their loss rate. This can be achieved by external magnetic perturbations. In TEXTOR it was possible to mitigate the runaway generation by the application of a resonant magnetic perturbation with toroidal mode numbers n = 1 and n = 2 using the dynamic ergodic divertor (DED). The disruptions were initi- ated by fast injection of about 3x1021 Argon atoms, which leads to a reliable generation of run- away electrons. At sufficiently high perturbation levels a reduction of the runaway current and, especially, the suppression of high energetic runaways with energies above 25 MeV was ob- served. These findings indicate the suppression of the runaway avalanche during disruptions, which is essential for ITER.
Fusion technology for and beyond ITER
ITER Diagnostics
Within 2008, the ITER core CXRS cluster of organisations, jointly led by FZJ and FOM/ITER- NL with participation of UKAEA, HAS and SCK.CEN, has performed comprehensive concep- tual design studies for the ITER core CXRS diagnostic system. As a part of these activities, FZK was focussing on the further development of the “reference option” for the components of the CXRS port plug. Specifically, a fast shutter concept was developed for protection of the
first mirror against deposition and erosion, together with a retractable tube, which allows for a frequent exchange of the first mirror via remote handling. Furthermore, concepts for the mirror holders and a shielding cassette were developed, to provide a robust and precise mounting of the mirrors as well as the necessary neutron shielding. In addition to these engineering activi- ties, neutron streaming and activation studies were performed, and the first mirror lifetime studies were continued.
By end 2008 FZJ received a substantial grant (11.6 M€) from the national Federal Ministry of Education and Research (BMBF) to perform R&D on ITER related projects, covering the pe- riod until the end of 2012. The major part of this budget is foreseen for the work on ITER core CXRS, with main emphasis on the development and the construction of prototypes for the port plug components, the development and testing of a prototype spectrometer (with ITER-NL) and the investigation of mirror degradation after neutron irradiation. Additional smaller sub- projects address the development of laser-based methods for the measurement of Tritium reten- tion in ITER, as well as the development of a fast valve technique for disruption mitigation at ITER.
Materials for components in contact with fusion plasmas
The institute IEF-2 operates the high heat flux test facilities JUDITH 1 and JUDITH 2. These electron beam facilities are capable for ITER- and DEMO-relevant quasi-stationary heat fluxes with loaded areas of up to 50 x 50 cm2 and transient thermal load tests in a millisecond time scale and energy densities in the MJ/m2 range to simulate Edge Localized Modes, plasma dis- ruptions, and vertical displacement events. A unique feature of this test equipment is the opera- tion inside a hot cell which allows testing of neutron irradiated and toxic materials (Be, T- containing samples).
The development, the qualification and the procurement of plasma-interactive components for next step devices such as ITER and DEMO are among the most challenging missions to be solved. Materials issues will play an important role to face the difficulties associated with the extreme environment in future fusion reactors. The particle fluxes will induce plasma-wall in- teraction processes which finally can degrade the materials with respect to their thermal and mechanical properties; in addition wall erosion is another critical issue which has significant impact on the lifetime of plasma facing components The plasma facing materials in the next step fusion device ITER are primarily based on beryllium, carbon and tungsten in combination with copper as a heat sink. In future fusion reactors such as DEMO and beyond refractory met- als such as tungsten or its alloys will play an even more important role.
A major approach of the R&D activities at FZJ is the characterization of plasma facing materi- als and actively cooled components and their assessment with respect to the thermo-mechanical behaviour under thermal transients (ELMs with an energy deposition up to 1.5 MJ/m2 and dis- ruptions up to approx. 5 MJ/m2). The research activities in 2008 are mainly oriented towards an improvement of multi-directional carbon-fibre-composites and monolithic tungsten as plasma facing materials for ITER and/or DEMO under the above mentioned transients. Special atten- tion was also paid to joints of these materials to CuCrZr and to thin tungsten coatings on a two- directional CFC for the ITER-like wall in JET under quasi-stationary loads (up to approx. 20 MW/m2). In addition, the participation in the European IP-project ExtreMat has contributed to
improve the database on future high temperature heat sink materials and alternative solutions for high heat flux components in DEMO such as Ti- or Si-doped carbon fibre composites and tungsten-fibre reinforced copper as a cooling structure.
Plasma-wall interaction
Plasma-wall interaction research is the main focus of the R&D activity of the fusion pro- gramme at Forschungszentrum Jülich. The research activities concentrate on the main and ur- gent critical ITER issues connected with the lifetime of plasma facing components and 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 mi- gration (lifetime), transient events (lifetime), material mixing (lifetime and safety), fuel reten- tion and removal (safety), high-Z plasma facing components (mainly lifetime) and dust (safety). At present two of these groups are led by FZJ scientists (material migration and tran- sients).
In line with the preparatory work for ITER is the contribution of FZJ to the currently largest tokamak JET within the Task Force E, where Jülich is strongly involved in experiments and in the task force leadership. In the JET-project ITER-like wall FZJ is responsible for the construc- tion of the full W-bulk divertor row and for various diagnostic enhancements.
The basic understanding of plasma-surface interaction and related plasma processes near plasma-facing components still needs to be improved. Modelling of material migration, thus erosion, transport and deposition in JET and TEXTOR, as well as the simulation of general plasma-wall interaction processes such as material mixing at the surface are an essential part of the work. The computer models have to be benchmarked with experiments and must be applied to provide reliable predictions for ITER.
R&D has been continued to qualify the chemical erosion of graphite under ITER-like low en- ergy detached plasma conditions. Recent dedicated JET experiments in high density L-mode discharges with a detached outer divertor again showed a large reduction of the chemically released carbon flux. In these experiments an in-situ calibration of the spectroscopy system by hydrocarbon injection was performed. The results show that low plasma temperature operation at the outer divertor in ITER (~ 2 eV) will lead to a significant reduction of the hydrocarbon flux, with the remaining carbon erosion dominated largely by the ELM phase where the tem- perature and ion flux is increased.
Dedicated experiments aiming at the characterisation of the plasma radiation under transient loads such as large Type I ELMs, disruptions and MARFEs have been performed in JET with the new bolometer system. These data are important for the estimation of the power loads on the wall in ITER with ELMs. The radiated plasma energy is proportional to the ELM energy at lost energies WELM below about 700 kJ. In this range the total ELM-induced radiation is about 50%
of the ELM energy drop, however with most of the energy being radiated after the fast ELM crash event. Beyond WELM of ~700 kJ, a non-linear increase of the divertor radiation has been observed which is interpreted as an additional carbon ejection from the target tiles due to abla- tion of the co-deposited layers in the inner divertor. In disruptions, the energy content in JET is
radiated in a poloidally asymmetric way with a peaking factor of about 2.5 at maximum for density limit disruptions. The peak radiation is located at the inner main chamber wall. This type of disruptions may elevate the temperature of the beryllium tiles in ITER close to the melting point (for the current quench alone).
The interplay between carbon and tungsten plasma facing materials has been studied in TEX- TOR by analysing the carbon deposition on bulk tungsten depending on the W-surface rough- ness, the substrate temperature and the carbon content of the W material. The surface rough- ness significantly enhances the C-deposition on W by an increase of the deposition layer thick- ness and an extension of the deposition area. Under constant plasma conditions, the C- deposition on tungsten vanishes at surface temperatures above about 500 °C. Also, the C- deposition rate increases when the concentration of C on the surface is increased by pre- implantation (up to 60% C) before plasma exposure.
Modelling
The modelling of long-term tritium retention in the divertor of ITER has again been carried out with the ERO code. Using a TriDyn-based surface model this time instead of the previously assumed homogenous material mixing, yields similar tritium retention rates. The estimations of tritium retention have been further improved by using surface temperature dependent tritium amounts in beryllium and carbon layers leading to a reduction of retention by a factor of about 2. The modelling of the impurity transport inside linear machines has been further improved and compared with experimental results from PISCES-B and Pilot-PSI. It has been seen that particle escape from the narrow plasma column is important for the interpretation of experi- mental observations. Also hydrocarbon transport modelling has been carried out for dedicated injection (CH4 and C2H4) experiments at TEXTOR. Modelled and measured penetration depths of CH, C2 and CII light and estimated D/XB values for CH agree well. Simulations of carbon deposition inside gaps of castellated surfaces have been continued and compared with experi- mental results from TEXTOR test limiter experiments. The observed deposition profile at the side surfaces of the gaps can be modelled if (at least) two carbon species with different sticking are considered.
ITER-like wall at JET
The ITER like wall project (ILW) has been launched in JET to study the compatibility of the ITER operating scenarios with metallic plasma-facing components (Be and W) and to address a number of dedicated outstanding PWI questions such as the long term Tritium retention un- der ITER material conditions. FZJ provides the leading Project Scientist (V. Philipps).
FZJ is responsible for the development of bulk tungsten divertor modules for the highly loaded divertor plates and for a number of diagnostic enhancements which are important to explore the specific issues of the ILW project (e.g. improved tungsten spectroscopy). The row of bulk tungsten tiles for the outer strike point in the divertor of JET is in the procurement phase and within the envisaged time schedule (March 2010). The remaining work has concentrated on the exact definition of the operational boundary conditions of the W bulk module.
Magnum-PSI
As part of the TEC collaboration the FOM-Institute for Plasma Physics Rijnhuizen, The Neth- erlands, is building a new machine to study plasma-wall interactions. The steady-state high particle flux density and a magnetic field of 3 T, and the large beam diameter will bring the relevant parameters typically an order of magnitude beyond what is presently available in lin- ear plasma devices, and thus into the realm of the ITER divertor.
The contributions of FZJ to this project are the concept, design and procurement of the target exchange and analysis chamber, which allows the surface characterisation and replacement of samples without breaking the vacuum of the main chamber. A significant part of the Magnum- PSI target exchange and analysis station has now been assembled. The design of the central carrier has been completed and is now under construction.
Theory and modelling
The main focus at FZJ in this area is the development and application of computational tools to quantify PWI related aspects in fusion edge plasmas, such as those resulting from wall released impurities, divertor chemistry and SOL turbulent transport. The new European supercomputer HPC-FF dedicated solely to fusion and hosted by FZJ, has lead to a further strengthening and focusing of this programme – with computational edge plasma science as one of its foreseen key application areas.
Integrated edge plasma modelling
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 and released (SOLPS4.3a). A distinguishing feature is the realization of Message-Passing (MPI) paralleliza- tion in the Monte-Carlo (EIRENE) part of the code. The code has meanwhile been tested on up to 1024 processors (within the EUFORIA-DEISA project at RZG). Full backward comparabil- ity with the ITER divertor design code SOLSP4.2 has been maintained. The upgraded code is in use jointly with the ITER team, in particular for the assessment of high divertor density con- ditions (ITER and DEMO), which have previously not been accessible due to run-time limita- tions. The large processor numbers and the nearly linear speedup of the EIRENE code also allow far more detailed numerical convergence studies of the coupled micro-macro system B2- EIRENE and a much higher level of numerical precision also in the previously accessible op- erational regime of the code.
A quantitative 3D analysis (EMC3-EIRENE) of integrated edge transport (plasma, radiation and neutral gas) in helical divertor configurations and under the influence of resonant magnetic perturbations (RMPs), as foreseen now also for ITER, has been carried out for TEXTOR-DED and D-IIID edge plasma conditions. Newly developed geometrical options (block structured grids) now also allow the application to poloidal divertor tokamak configurations (e.g. DIII-D, JET, MAST, ASDEX-U) and first comparative results with regard to TEXTOR-DED and DIII- D poloidal divertor configurations have been successfully carried out.
Impurity transport
The behaviour of impurities and their impact on fusion plasmas is of key importance for pre- sent and future devices. The physical origin of the plasma confinement modifications induced by deliberate seeding of impurities into JET is investigated by using the RITM code in order to identify responsible physical mechanisms. It is found that as for the transition to the RI-mode in TEXTOR, the suppression of ITG turbulence by the increased effective ion charge is of most importance for JET conditions. The model for the anomalous impurity transport used in RITM has been elaborated further by including the effects from coulomb collisions between impuri- ties and the main plasma particles. Calculations for JET indicate that collisional effects such as friction and thermal forces provide new channels for the anomalous convection and can, in particular, explain its dependence on the impurity ion charge. New mapping methods devel- oped for the description of the field line behaviour in the presence of RMPs will be applied to study the impurity transport in stochastic magnetic fields.
European Transport Solver within the EFDA ITM Task Force
The main objective of this activity is to provide the computational basis for a modular transport code – the European Transport Solver (ETS) – aiming at a self-consistent modelling of plasma parameters in the core, the pedestal and the scrape-off layer – and ultimately at the simulation of complete discharge scenarios, e.g. for ITER. For this, transport codes being used in Euro- pean Associations have been reviewed, the physics content and the numerical concept of ETS have been defined and a standardized data structure for 1-D transport computations has been formulated. In particular, new numerical approaches to integrate the highly non-linear transport equations into the codes RITM and ETS have been further developed.
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.
The 3D drift fluid turbulence code ATTEMPT (FZJ) has been extended to take into account modifications of the magnetic confinement field on large and small scales self-consistently.
Additionally, the code has been prepared to run with good scalability on multi-processor- platforms allowing for studying the computationally demanding interaction of large and small scales with proper resolution. As a key finding a bifurcation mechanism could be identified, governing the interplay of plasma currents and electric fields, i.e. plasma rotation. This gives a strong indication that this bifurcation mechanism explains the sudden change of plasma con- finement and plasma rotation in the presence of RMP. The dependence on collisionality found in the numerical simulations might give a hint for the controlled use of RMPs for ITER rele- vant operational windows.
Nuclear Fusion Programme – Progress Report 2008 B.1. Plasma-Wall Interaction
Volker Philipps (IEF-4 Plasma Physics, v.philipps@fz-juelich.de)
Introduction
Solutions of remaining difficulties in operation of the next-step fusion device ITER, especially those related to plasma-wall interaction are urgently needed in the current phase in order to make the best engineering choices to ensure a safe and reliable plasma operation. Operational experiences must be gained in parallel to the ITER construction for a number of additional spe- cific topics to support a safe ITER operation. The Main Topic Group on Plasma-Wall Interac- tion (PWI) remains organised in a programmatic oriented fashion and with the motivation of developing a viable solution to the questions of the ITER first wall and divertor materials as well as to possible operation scenarios to reduce power and head loads to the first wall and in particular the divertor. In this sense, the Main Topic Group is fully aligned with the goals of
the European Task-Force on PWI see ITER-relevant
subjects are thus treated with priority.
In line with the preparatory work for ITER is the strong contribution of the Main Topic Group to the currently largest tokamak JET, within the associated Task Force Exhaust (see http://users.jet.efda.org/pages/e-task-force/index.html.). At JET, ITER like wall material com- binations will be tested in the frame of the ITER-like wall project acting as a test bed for the currently foreseen material options in ITER. In parallel, the basic understanding of plasma- surface interaction and related plasma processes near plasma-facing components is still to be improved. Modelling of material migration, thus erosion, transport and deposition in JET and TEXTOR, as well as simulation of general plasma-wall interaction processes such as material mixing at the surface are essential part of the work of the Main-Topic Group. The applied modelling is benchmarked with experiments and applied to provide predictions for ITER for the key PWI questions for ITER operation.
The two categories of open questions for PWI on ITER are clearly: (i) safety issues (operation) with respect to tritium inventory and (ii) the lifetime of the plasma-facing components (PFC) such as divertor target plates. (i) includes the measurement and the prediction of fuel retention via reliable modelling, as well as techniques to clean up the plasma-facing components and control the dust inventory. (ii) is focused on the modification of plasma-facing material, and here especially of the target and other highly exposed elements, with long-time operation, thus mainly related to erosion, sublimation and melting. The need for qualification of high-Z mate- rials such as W (or others) which are suitable for plasma-facing components has became a re- search focus of PWI in the last year.
Main Topic on Plasma-Wall Interaction deals therefore in first instance with the following fields:
(a) Carbon-based PFCs: Erosion, transport and deposition, and thus material migration and tritium retention. Here, ITER components made on a carbon basis are addressed. The de- velopment of deposition mitigation techniques and of removal methods for hydrogen iso- topes belongs to this field. Work concentrates on erosion at ITER-like divertor conditions, transport of carbon along surfaces and to gaps and qualification of fuel removal techniques in the presence of magnetic fields
(b) High-Z PFCs: Tungsten is foreseen for the baffles or – during the activated phase in ITER also for the divertor plates. Research in this field concentrates on high temperature behaviour of W close to and at the melting and on retention of hydrogen in the bulk material.
(c) Material mixing: The ITER first wall material mix may lead to material mixtures on PFCs and to the appearance of alloys, carbides, etc. These effects are crucial for material properties and fuel retention. Research concentrates on mixing of W with C by co-deposition and im- plantation.
(d) Plasma operation under detached conditions: The operational divertor regime in ITER requires divertor detachment to avoid damage of the divertor target plates. This high den- sity operation is close to the appearance of radiative instabilities (MARFE) which finally can lead to disruptions.
(e) Qualification of atomic and molecular data: Modelling of plasma-wall interaction proc- esses as well as optical edge diagnostics rely on a good atomic and molecular database.
The Plasma-Wall Interaction group mainly works on TEXTOR and JET with additional contri- butions from ASDEX-Upgrade (AUG), DIII-D, Tore Supra, PISCES, Pilot-PSI, and other fa- cilities. The full research programme is organised within the Trilateral Euregio Cluster (TEC).
The partners Japan, USA and Canada within the "IEA Implementing Agreement on Plasma- Wall Interaction in TEXTOR" are closely linked to the research programme, as shown in this report in section C. The TEXTOR tokamak serves as the central fusion facility for the TEC partners, without prejudice to resorting to any other device if better suited. Joint experiments are performed at different machines in the frame of ITPA Divertor Scrape-Off Layer working groups. These working groups directly address requests from the ITER PWI science commu- nity (EFDA, ITER Organisation, F4E, etc).
1. Material erosion, migration and deposition
1.1. Chemical erosion, hydrocarbon catabolism and spectroscopy
Carbon-based materials such as carbon-fibre components (CFC) have been deployed success- fully as plasma-facing materials at the locations of highest heat flux, such as the divertor target plates in present-day fusion devices. Therefore, the current ITER material choice foresees CFC at the divertor areas with highest heat flux for the non-activated operation phase. However, chemical sputtering of carbon and the appearance of fuel co-deposition are critical issues in the usage of carbon with respect to machine reliability, and in the case of tritium also with respect to machine safety. It is important to note that carbon erosion – with the foreseen operation time of one full set of divertor modules and with the predicted number of moderate ELMs – is no longer considered as a lifetime issue for the ITER wall.
The main question is if the remaining CFC will be acceptable for a short operation period dur- ing the activated phase of ITER. In case it is acceptable, this would allow to demonstrate Q =
420.0 422.0
/ nm
CD A-X band+ C II
modelling (K.Behringer)
intensity / arb. units
-2.0
Trot2000 K
TEXTOR
#98026
424.0 426.0
0.0 2.0 4.0 6.0 8.0
10 quite early in the operational phase of ITER. This question is closely related to the determi- nation of the sputtering yields under ITER-relevant plasma conditions because a substantial part of the tritium inventory is found in the carbon-based re-deposited layers. Also, this point is related to the production of mixed materials, i.e. the impact of Be in the deposits, and the pos- sibilities to clean in-situ the locations with large deposits and remove incorporated fuel.
Carbon is sputtered at the target plates and is then subsequently migrating to remote areas.
Quantification of CFC erosion yields and their estimation with erosion and deposition simula- tion codes, e.g. ERO developed and maintained at FZJ (see below), is a topic of the ITPA Di- vertor and SOL Group and of the SEWG (Special Expert Working Group) on Material Migra- tion in the EU Task Force on Plasma-Wall Interaction. Experiments under the scientific leader- ship of FZJ have been performed in 2008 at JET and TEXTOR, as well as in the linear ma- chine Pilot-PSI at FOM.
Experiments in 2008 mainly focused on
a) the data base (HYDKIN) and code validation (ERO-code) for higher hydrocarbons,
b) the measurement of erosion yields or conversion factors in situ, with the aid of optical spec- troscopy on the CD radical and local injection of hydrocarbon species, and
c) the measurement of erosion yields at low electron temperature and high density plasmas.
1.2. Comparison of measured and modelled effective D/XB values from hydro- carbon injection experiments
At TEXTOR, a series of experiments has been carried out to determine the inverse photon effi- ciencies, in particular of the pair CH/Gerö band, thus, to verify the HYDKIN database and the plasma-related part of the ERO code. These experiments were performed with gas injection modules made of metal and a minimised contact surface area to exclude any surface effects.
Experimental results, i.e. photon efficiencies for typical ohmic TEXTOR plasmas with electron temperatures of about 50 eV at the LCFS as well as HYDKIN calculations were presented in the 2007 report. In 2008 the focus was set on the comparison with ERO for higher hydrocarbon species (described below), the production of CD+ and the impact of local cooling due to the local injection. The comparison with ERO for higher hydrocarbons seems to be in good agree- ment with respect to the production of C2 and the corresponding effective D/XB values for the d-a Swan band.
Fig. 1: Douglas-Herberg band at 420 nm in overlap with the CD Gerö-band .The spectrum was taken from the difference between an ethane and methane injection
(Brezinsek et al.: Journal of Nuclear Materials, 2007).
Fig. 1 describes the CD+ band identified in the experiments with methane and ethane injection.
The CD+ Douglas Herzberg band seems to be preferably detectable with the injection of higher hydrocarbons. The quantification with simulation is ongoing with the aim to provide D/XB values for CD+.
1.3. Reduction of the hydrocarbon flux in the outer divertor of JET during diver- tor detachment
High density L-mode discharges in JET with outer divertor detachment have shown a reduction of the CD A-X band light which is an indicator of chemical erosion of carbon-based plasma- facing materials. The strong reduction in CD emission and the moderate reduction in CII can be clearly seen in fig. 2a which shows the integrated light emission over the outer divertor tar- get plate during a density ramp-up discharge with MARFE formation. Spatially resolved in- formation showed that the detachment started at the outer strike-point and extended during the density rise over the full target plate (DOD ~3 for the target plate). The remaining emission in CD and CII is a result of emission from the vertical target where the ionisation front has been shifted up.
Fig. 2: a) Time traces of CII, CD and Dg line intensities during a density ramp-up discharge approach- ing detachment. b) Injection of CD4 leads to a strong increase of the CD band which is not detectable
otherwise in detached conditions.
An in-situ calibration with hydrocarbons was performed in order to distinguish the different roles of a change of the photon flux and the particle flux. CD4 was injected through GIM 14, a single valve in the load bearing septum replacement plate or outer target plate, into the near scrape-off layer of the outer divertor. The target plasma was a high density L-mode plasma with density feedback control. The injected amount of molecules leads to an emission of ex- trinsic CD photons indicating that the reduction of the intrinsic CD emission is related to a re- duction of the carbon flux. The corresponding effective D/XB value has been determined to about 45 in the detached phase at ne ~ 2x1020m-3 and Te below 2 eV and to about 30 in the at- tached phase. Plasma parameters have been determined in-situ by Balmer spectroscopy and
Langmuir Probes measurements. This gives clear evidence for a reduction of the chemical ero- sion of CFC in the phase with detached outer divertor plasma. However, the strong reduction of the hydrocarbon particle flux of more than a factor 3 is partially compensated by the reduc- tion of the impinging ion flux when the divertor detaches. The erosion yield itself is only sig- nificantly reduced when the particle flux of neutral atoms to the target is taken into account.
These experimental results confirm the Roth formula for detached plasma conditions in L- mode if both impinging fluxes, the atomic flux and ion flux, are considered. (Brezinsek et al.
PSI 2008 / JNM 2009).
A second series of experiments was done in high density H-mode discharges with partial de- tachment of the outer divertor leg. The high density operation caused a type III ELMy H-mode regime. A reduction similar to the L-mode case in-between ELMs could be observed. How- ever, the energy drop per ELM was still large enough to re-attach and to create a hotter plasma edge causing again physical and chemical sputtering. Therefore, the suppression was not as strong as in the L-mode case. Local hydrocarbon injection has been performed but results have not been fully analysed yet and are subject to calibration of spectroscopic systems at JET.
The consequences for ITER can be described in the following way:
i) Low temperature plasma operation at the outer divertor (below 2 eV) leads to a reduc- tion of the hydrocarbon flux – being the indication of an energy threshold?
ii) Remaining carbon erosion is dominated by the ELM phase where the temperature and the ion flux are increased.
iii) The overall erosion of graphite at the outer target plate is reduced.
1.4. Impact of N2 as a seeding gas on the chemical erosion in the JET divertor Apart from strong deuterium injection, also combined impurity seeding and deuterium puffing have been applied to reach outer divertor detachment and eventual recombining plasma condi- tions in the outer divertor leg. Surprisingly, the additional injection of N2 leads to a decrease of both CH and CII line radiation, which represents chemical and total sputtering. A reduction of the local temperature might be a possible explanation for the reduction which would be in line with the previously described observation of reduction of chemical sputtering with pure deute- rium puffing. On the other hand, the signature of nitrogen chemistry is clearly detectable – CN has been measured and identified. CN might either be produced at the surface by nitrogen sput- tering or it can be generated in the plasma as a by-product of the CH scavenger by nitrogen.
Figure 3 shows the CN spectrum and the time trace during a discharge with nitrogen seeding with respect to a discharge without. Please note that the Balmer-lines are increased which indi- cates a higher density and a colder local plasma.
a)
b)
Fig. 3: a ) CN emission spectrum during the circumferential injection of N2 in the outer divertor plasma. b) Time traces of CII and CD during two comparable plasma discharges with and
without N2 seeding. The reduction of the emission integrated over the full outer divertor leg is clearly visible.
1.5. Material migration studied by means of a Quartz Microbalance (QMB) in JET and TEXTOR
QMB operation in JET
After an extensive repair of the QMB-systems in the 2007 JET shut down QMB measurements were continued in 2008 resulting in two ITER relevant publications [1,2]. Material deposition was measured in the private flux region of JET with QMB for ~ 11,900 s of successive plasma discharges with moderate additional heating (2–3 MW). The QMB was located under the load bearing septum replacement plate with a view towards the inner divertor. In total the QMB detected material deposition equivalent to a ~ 270 nm thick hydrogenated amorphous carbon layer (assuming a density of 1 g/cm3). This appeared for an integrated particle fluence of 2.3×1023 D+ ions into a toroidal section of 1 cm length of the inner divertor. The area of the QMB in the private flux region was deposition dominated when the inner strike point position was on the vertical tiles, line-of-sight with the quartz crystal, and turned into erosion solely by moving the strike point from vertical tile 3 to horizontal tile 4. The most likely reason is a
1 Erosion and deposition behaviour of a-C:H layers in the private flux region of the JET MKII-HD divertor H.G. Esser, A. Kreter, V. Philipps, A.M. Widdowson et al., Journal of Nuclear Materials, Volumes 390-391, 15 June 2009, Pages 148-151
2 A. Kreter, H. G. Esser S. Brezinsek, J. P. Coad et al., PHYSICAL REVIEW LETTERS, PRL 102, 045007 (2009)
change of the C/D flux ratio of the particles impinging on QMB turning this area from a depo- sition to an erosion dominated region.
The impact of edge localized modes (ELMs) carrying energies of up to 450 kJ on carbon ero- sion in the JET inner divertor has been assessed with the QMB system positioned near the lou- vers in the inner divertor area. It has been found that the inner target erosion is strongly nonlinearly dependent on the ELM energy: a single 400 kJ ELM produces the same carbon erosion as ten 150 kJ events. The ELM-induced enhanced erosion is attributed to the presence of co-deposited carbon-deuterium layers on the inner divertor target, which are thermally de- composed under the impact of ELMs.
1.6. Material migration and redeposition studies with a QMB in TEXTOR
Fig. 4: Frequency change of the QMB-crystal depending on the total amount of injected CD4
through a hole next to the crystal.
Fig. 5: Frequency change of the QMB-crystal depending on radial position under constant CD4 injection conditions.
A first QMB local migration experiment with a newly developed limiter head was carried out in TEXTOR. The head was inserted into the TEXTOR material test facility which allows posi- tioning of the QMB-crystal at different radial positions. CD4 was injected through a hole in the
vicinity of the QMB-crystal and the layer growth was measured in situ and shot resolved de- pending on the radial position and the total amount of injected CD4.
As an example figure 4 shows the frequency change of the QMB-crystal corresponding to the mass change of the layer on top of the crystal as function CD4-injection under identical plasma conditions. A clear linear dependence is found. Figure 5 shows the dependence of the layer growth as a function the radial position with respect to the LCFS with constant amount of CD4- injection. ERO modelling of these data is planned for the future.
1.7. Material migration and fuel retention studies in ITER-like castellated structures in TEXTOR and DIII-D
Since the entire first wall and divertor armour of ITER will be castellated fuel accumulation in the gaps of castellated structures can lead to undesirable tritium inventory limiting the opera- tion time of ITER and representing a safety issue. Measures must be taken to reduce the impu- rity deposition and fuel inventory in the gaps. Shaping of the castellated cells is a possible strategy to mitigate deposition in the gaps.
A castellated limiter made from tungsten carrying rectangular and roof-like shaped cells was exposed in the SOL plasma of TEXTOR to series of reproducible well-diagnosed neutral beam-heated discharges [3]. Cells of different shapes were exposed under similar plasma con- ditions allowing for direct comparison of deposition patterns. After exposure, an absolute quantification of carbon and deuterium amount was made for both toroidal and poloidal gaps of shaped and rectangular cells by means of several ion-beam and electron-beam surface diag- nostics. A factor of 3 less deuterium was found in the gaps of shaped poloidal cells compared with rectangular ones. The difference in the carbon deposition was less pronounced, calling for a further optimization of shaping. Toroidal gaps contained comparable amounts of carbon and deuterium and cannot be ignored when accounting the fuel inventory and impurity transport.
Significant tungsten intermixing was observed reaching up to 70 at. % of W in the carbon de- posited layers in the poloidal gaps. Metal intermixing will significantly decrease the efficiency of cleaning of gaps in ITER. Gaps contained about 10% of the total amount of carbon and less than 0.01% of deuterium impinging the castellation. A low trapping ratio for D may be caused by temperature excursions during exposures.
Based on these results a new experiment with optimized shaping of castellation was made in DIII-D in the frame of a new collaboration between TEXTOR and DIII-D tokamaks in this area. A castellated sample was placed in the DiMES material transport system and exposed to 10 highly reproducible ELMy H-Mode discharges in the private flux region (PFR) of DIII-D divertor (Figure 6). After exposure thin stripe-like deposition patterns were detected in the gaps of exposed cells similarly to observations made on gaps of a castellation exposed in TEXTOR.
Surface analysis is ongoing.
3 A. Litnovsky et al., “Investigations of castellated structures for ITER: the effect of castellation shaping and alignment on fuel retention and impurity deposition in gaps”, presented at the 18th Int. conference on plasma-wall interactions in controlled fusion devices (PSI-18), Spain, May 2008, accepted for publication in J. Nucl. Mater.
PFR
Tungsten castellated Di- MES sample
Plasma flow
3 geometries of cells
Moderate
shaping Strong shaping No shaping (rec-
tangular geome- try)
This research is part of the multi-machine IEA-ITPA joint experimental programme, task DSOL 13.
Fig. 6: Scheme of the new experiment with castellated structure in the DIII-D tokamak: geometry of exposure, photograph of the castellated DiMES sample and 3 different geometries of exposed cells.
2. Fuel retention and removal
2.1. Fuel retention measurement by laser induced desorption spectroscopy Long term tritium retention is one of the most critical issues for ITER and future fusion de- vices. To identify the mechanism, location and amount of retention, its dependence on plasma and wall conditions and to qualify T retention mitigation and control techniques, laser induced desorption spectroscopy of retained fuel has been developed and applied in TEXTOR. Hydro- gen isotopes are desorbed from re-deposited layers by rapid heating with laser radiation and the released particles have been quantified in situ by spectroscopic measurements of hydrogen lines in the TEXTOR plasma.
With this in situ method the hydrogen inventory in deposited hydrocarbon layers was measured reliably in TEXTOR. The detection limit was about 2x1020/m2. This is about the value, which was detected also in the erosion dominated area indicating a dynamic equilibrium. The lower limit is set by the natural fluctuations of the hydrogen background light. Assuming a density of 2x1019/m2nm H atoms in hydrocarbon layers the detection limit corresponds to a 10 nm thick layer.
2.2. Fuel retention and erosion under ITER-like mixed species plasma conditions in PISCES
Samples of CFC NB41 and fine-grain graphite ATJ 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 linear plasma device PISCES 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 deuterium retained 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 ( = 11025 – 51026m-2), ion energy (Ei = 20 – 120 eV) and sample surface temperature (Ts = 370 – 820 K). It has been found, that for Ts = 470 K the retention scales as 0.35. For Ts = 820 K the retention saturates at a level of ~1021 D/m2. The retention increases with Ei and drops with higher Ts. At Ts = 720 K, the beryllium seeding results in a building of a protective beryllium carbide layer, which appears to prevent the in-bulk diffusion of deuterium, thus reducing the retention. Admixture of Ar and, in the case of low Ei, He leads to a significant reduction of the retention. Also, the addition of Ar and He has not resulted in an increase of the carbon erosion rate. Moreover, the characteristic time for the Be-layer formation was in agreement with the scaling [D. Nishijima et al, J. Nucl. Ma- ter. 363–365 (2007) 1261] within its uncertainty. It can be concluded, that for the covered range of experimental parameters the addition of Ar and He to plasma does not affect the build- up of the protective Be-carbide layer and associated mitigation of carbon erosion.
2.3. Fuel release after JET disruptions
The amount and temporal behaviour of the fuel release from the JET walls has been analysed for normal and disruptive shots for the previous JET SRP divertor campaign (2002-2004, about 8000 shots). For all non disruptive shots, the fuel release is determined by the dynamic trap- ping of fuel in the plasma loaded graphitic surfaces. The total dynamic hydrogen inventory is ~ 3x1022 corresponding to about 2x10 20D/m2 if uniformly distributed over the JET wall area.
The dynamic inventory depends only weakly on the total amount of fuelling/shot indicating that this dynamic inventory is close to saturation in JET and could not be increased much if JET would e.g. operate with longer plasma pulses (higher wall particle fluencies). After the shots, the dynamically retained fuel is released with a quite uniform power law tn law with an exponent n = -0.6. In disruptive shots additional fuel is released which increases linearly with the stored energy at the time of disruption. The averaged fuel release is about twice that for normal shots but maximum values at the highest stored energies up to 5 times the normal re- lease are observed. Fig 7 shows the particle release for all disruptions in the database as a func- tion of the stored energy just before the disruptions. The temporal behaviour of the disruptive fuel release follows largely the time decay of the vacuum pump down time showing that the fuel is released instantaneously at the disruption. The most likely process for this is thermal desorption from overheated wall areas due to the disruptive power loss to the walls. If the aver- age additional release of ~ 2x1022atoms/disruption is attributed only to the thermal release from codeposits, and assuming a disruption frequency of 10%, a total additional fuel release of 30g can be estimated per typical JET campaign. This is in the range of the campaign averaged re- tention observed by post mortem analysis, e.g. 60 g in the SRP divertor campaign. This process
0.0 2.0x106 4.0x106 6.0x106 8.0x106 0.0
2.0x1022 4.0x1022 6.0x1022 8.0x1022 1.0x1023
Stored energy (J)
Integrated release after shot
Mean value
can contribute to the observed fraction of fuel retention on the plasma wetted deposits in JET which is typically 0.1-0.2 D/C+Be but reaches higher values on plasma hidden areas (0.5-1 D/C ).
Fig. 7: Particle release integrated up to 700 s after the shot for all disruptive shots.
3. Fuel removal and wall conditioning
Ion Cyclotron Wall Conditioning (ICWC) and normal GDC has been continued in TEXTOR and a large toroidal simulation device (Tomas facility). GDC cannot be applied in the presence of magnetic fields and even the residual magnetic field in ITER due to the magnetic inserts may represent a concern for the application of GDC when the field is turned off. The toroidal and poloidal homogeneity of GDC has been studied in Tomas. Fig. 8 shows the relative GDC current in dependence of an externally applied magnetic field on different positions with a GDC operating from one anode fixed to one position (simulating tokamak conditions). Already at about 4 mT, the GDC current at positions away from the antenna vanishes while the current concentrates more and more to the vicinity of the anode. This restricts GDC wall conditioning in ITER to the periods with Bt off and calls urgently for development of alternative condition- ing methods which can be applied in the presence of magnetic fields.
Fig. 8: Local GDC current in a toroidal geometry with one fixed GDC antenna depending on external magnetic field.
0 50 100 0
5 10
ICRF plasma
Ar removing rate [1016 atom/s]
time [s]
He+H2 #107208 He+H2 #107218 He+NH3 #107221 D2+NH3 #107224 D2+N2 #107227
Fig. 9: Ar removal rate for ICWC plasma application in different gas mixtures in TEXTOR. The TEXTOR walls were preloaded before each ICWC plasma with Ar in a GDC plasma in Ar/D2. Wall cleaning has been further studied in ICWC plasmas in various gas mixtures, magnetic field configurations with overlaid radial and vertical fields and RF operation modes [4,5,6].
Fig 9 shows the removal of Ar that has been previously loaded to the TEXTOR walls by GDC in a D2-Ar mixture following ICWC wall cleaning in different types of gas mixtures.
4. Transient heat loads
Runaway dynamics and wall interaction in JET. The limits for the energy impact during transients can be deduced from the one-dimensional solution of the heat diffusion equation in a semi-infinite solid. These limits are defined by the melting/sublimation temperature of the PFC material and are 15 MWm-2s0.5 for beryllium and about 40-60 MWm-2s0.5 for graphite and tungsten. These estimates are valid, if the heat is deposited at the surface of the PFC. However, because of the high energies in the MeV range, the runaways have a non-negligible penetration depth. Assuming an exponential radial decay of the energy deposition, an analytical formula for the temperature increase can be found:
where = K/c, K is the heat conductivity, c the heat capacity, and the density. is the radial e-folding length of the heat source and q is the heat flux density. The deposition duration can be estimated from the final decay of the runaway current: loss = IRE(dIRE/dt)-1. This value varies between 2 ms and 5 ms for JET runaway disruptions. The penetration depth in Be and C is 2.5 mm and 2.0 mm, respectively. The penetration in W is 0.15 mm. A mean energy of Eav = 12.5
4V. Philippset al 1), development of wall conditioning and tritium removal techniques in TEXTOR for ITER and future fusion devices, IAEA conference Geneva, 2008
5 A. Lyssoivan, influence of toroidal and vertical magnetic fields on Ion Cyclotron , Wall Conditioning in tokamaks, Journal of Nuclear Materials 390–391 (2009) 907–910
6 G. Sergienko et al, ion cyclotron wall conditioning in reactive gases on TEXTOR, Journal of Nuclear Materials 390–391 (2009) 907–910
MeV is assumed, which is expected after avalanching. Runaways generated by the Dreicer mechanism gain even higher energies and would penetrate deeper. The resulting critical energy densities are: 6 MJ m-2 for Be, 3-5 MJm-2 for W, and 11-13 MJ m-2 for C. These values are significantly larger compared to those for surface deposition: 6-8.5 times for Be, 4.5-6.5 times for C. Because of the shallow penetration, they remain almost unchanged for W.
The heat load generated by runaways is discussed in the following for the JET pulse 68782.
During this disruption a runaway current of 0.5 MA was generated. A runaway plateau of about 5 ms is formed and the runaway beam is finally lost to the upper dump plate. The dy- namics of the runaway beam can be detected by the radiation from K-shell vacancy production, which is recorded by the soft X-ray camera (figure 10). The runaway beam builds up at the vertical centre and moves then with a velocity of about 200 m/s upwards while increasing in size. Finally, the beam touches the upper dump plate and releases its energy to the graphite tiles. With a beam radius of about 0.5 m, a heat deposition time of 2.5 ms is estimated. This value is in agreement with the current decay time of 3 ms.
The impact of the runaway beam is detected by the wide angle IR camera (figure 11). Two frames have been recorded during the disruption. The first frame shows a strong temperature increase in the divertor and at the inner poloidal limiter caused by the heat flux during the en- ergy quench. However, on the second frame recorded 20 ms later, a temperature rise at the up- per dump plate is visible, which can be attributed to the impact of the runaway beam.
Fig. 10: Soft X ray radiation during the current quench for shot 68782.
A detail of the upper plate is seen in figure 11b, giving the temperature difference between the two frames. The maximum temperature rise is about 530 K. The heat is deposited on a very small area of about 0.3 m2. The inhomogeneous load might result from small misalignments of the tiles, which become significant at shallow angles of incidence.
The heat flux density can be estimated from the above values for runaway current, loss time and runaway energy of 12.5 MeV, to be q = 400 MWm-2. The assumption for the energy is supported by the observed neutron production, which has a threshold energy of about 10 MeV.
However, the detailed energy spectrum is unknown. According to the above equation, this heat
load leads to an temperature rise of T 330 K, well in agreement with the measurements. For surface deposition the temperature rise would have been T ~ 1500 K.
(a)
(b)
Fig. 11: Heat load distribution during the disruption 68782, a) overexposed frame for visualisation, b) temperature rise at the upper dump plate due to runaway impact.
Fig. 12: Radiation distribution during the disruption.
