Fusion Plasma Physics Annual Report Division of Fusion Plasma Physics Alfvén Laboratory School of Electrical Engineering KTH

Transcription

1 Fusion Plasma Physics Annual Report 2006 Division of Fusion Plasma Physics Alfvén Laboratory School of Electrical Engineering KTH Stockholm, May 2007

2 Division of Fusion Plasma Physics, Alfvén Laboratory, School of Electrical Engineering, KTH, Stockholm, Sweden 2

3 Contents 1. Executive summary Overview EXTRAP T2R experiment Active MHD mode control system EXTRAP T2R enhancements during Summary of research results Resistive wall mode control Intelligent shell feedback Linear coupling of side band modes Fake rotating shell feedback Comparison of different magnetic field sensors Mode control with complex feedback gain Non-linear MHD dynamics and plasma turbulence Tearing mode dynamics Plasma turbulence, flow and transport Plasma-wall interactions Studies of fuel retention and material migration Development of Be coatings for the ITER-Like Wall Project Deposition and fuel inventory in C and Be MkI divertors First mirror test at JET for ITER Material mixing on tungsten limiters Fuel removal from PFC by laser irradiation Plasma diagnostics Theoretical fusion plasma physics Fast particle excitation of global Alfvén eigenmodes Minority ion cyclotron current drive in ITER Ion cyclotron emission in toroidal plasmas RF-induced rotation Ripple experiments Ion cyclotron current drive for monster sawtooth control in JET Anisotropy drive for geodesic acoustic and tornado modes Production of fast deuteron tail by ICRH accelerated He 3 ions EFDA-JET activity Computational methods for fusion plasmas Confinement studies of the advanced RFP Semi-analytical solution of initial-value problems Chaos and self-organization International collaborations Education and research training Undergraduate education Graduate education Publications Peer-reviewed journal publications International conference contributions Staff Professional activity

4 9.1. Membership, honours, responsibilities Academic and expert activity Economy

5 1. Executive summary Research areas Research carried out at the Division of Fusion Plasma Physics is part of the European magnetic fusion programme. The Division is hosting one of the specialized fusion experiments in Europe, the EXTRAP T2R device. Research at the Division is presently focussed on the following topics: Resistive wall mode control. Non-linear MHD dynamics and plasma turbulence. Plasma-wall interaction. Plasma diagnostics. Theoretical fusion plasma physics. Computational methods for fusion plasmas. Chaos and self-organization. Research results 2006 Many interesting research results have been obtained during 2006: Studies of feedback methods for optimization of resistive wall mode active control. Tearing mode dynamics from measurements of mode velocity with magnetic coil arrays and spectroscopic measurements of plasma flow. Investigation of the relationship between turbulence and the mean flow using probe measurements with high spatial and temporal resolution. Studies of material erosion, migration and re-deposition. Fuel retention studies and fuel removal techniques. Characterization of plasma-facing materials testing high-z metals. Development and characterization of wall materials for the Iter-like wall project at EFDA-JET. Development of diagnostic tools for plasma wall interaction studies. Simulation of the dynamics of TAE excitation during ICRH. The effect of the orbit topology on ion cyclotron emission. The effect of ICRH on the toroidal rotation of the plasma. Fast ion losses and transport of thermal ions due to toroidal field ripple. Destabilization of sawteeth by minority ion cyclotron current drive. Numerical study of stability and confinement in the advanced RFP. The role of electric field for transport barriers. 5

6 Publication statistics The research is very productive as is evidenced by a record-high number of scientific publications for Researchers at the Division of Fusion Plasma Physics have during 2006 published 44 articles in peer-reviewed journals and contributed 58 papers to international conferences and workshops. International collaboration The Division of Fusion Plasmas Physics participates in the work programme at EFDA-JET in the areas of ion-cyclotron resonance heating, plasma-wall interaction, spectroscopy diagnostics, and is starting activity on magnetohydrodynamic instabilities. The Division co-ordinates projects in the two European Task Forces on Integrated Tokamak Modelling and Plasma-Wall Interaction. It contributes to ITER within the EFDA Technology programme. Collaborations with other European fusion associations include TEXTOR, ASDEX-U and RFX-Mod fusion research groups. Graduate education Two graduate students at the Division of Fusion Plasma Physics were awarded PhD degrees during 2006: Dmitriy Yadikin, Ph D degree for the thesis "Resistive wall mode stability and control in the RFP". Jon-Erik Dahlin, Ph D degree for the thesis "Numerical studies of current profile control in the RFP". In addition, one graduate student at the School of Engineering Sciences, Department of Physics was awarded a PhD degree for work carried out at the EXTRAP T2R device: Mattias Kuldkepp, Ph D degree for the thesis "Diagnostics for advanced fusion plasma scenarios". Undergraduate education The Division of Fusion Plasma Physics participates in the Erasmus Mundus European Master Programme in Nuclear Fusion Science and Engineering Physics (FUSION- EP). The programme started in Other participants are Ghent University (Gent), Universidad Carlos III (Madrid), Universidad Complutense de Madrid, Université Henri Poincaré (Nancy), and Universität Stuttgart. Development work was initiated for a new KTH Master of Science Programme in Electrophysics, as a collaboration between the divisions of Fusion Plasma Physics, Space and Plasma Physics and Electromagnetic Engineering. The application is to be finalized during spring

7 2. Overview Developing fusion as a new clean, safe and virtually inexhaustible energy source is the ultimate goal of fusion research. This is one of the world s most demanding scientific challenges, and has lead to the construction of large scale experimental facilities. The next step is ITER, a world wide collaboration, in which Europe takes a leading part by hosting the facility in Cadarache, France. The aim of ITER is to demonstrate the scientific and technological feasibility of a fusion reactor. European Fusion Programme The European magnetic fusion research programme is based on the Euratom treaty. Participation in the programme by national research organisations is co-ordinated through Contracts of Associations. The EU funding to the Division of Fusion Plasma Physics is based on cost sharing at various percentages where support from Swedish sources is necessary. Research at the Division of Fusion Plasma Physics Decisions on activities are made by a Steering Committee with representation from the European Commission. The research at the Division includes projects on Resistive wall mode control. Non-linear MHD dynamics and plasma turbulence. Plasma-wall interaction. Plasma diagnostics. Theoretical fusion plasma physics. Computational methods for fusion plasmas. Chaos and self-organization. EXTRAP T2R The Division is hosting one of the smaller specialized fusion experiments in Europe, the EXTRAP T2R device. The experiment has special capabilities for studies of active MHD instability control. EXTRAP T2R was upgraded during in collaboration with Consorzio RFX, Padova, Italy. A comprehensive system for feedback control of MHD modes was developed and installed at EXTRAP T2R. The Active MHD Mode Control System has capabilities that at present are un-paralleled in any tokamak fusion device, and the device provides excellent capabilities for research on active control of MHD instabilities, a research field which has high priority in the European fusion programme. EFDA-JET The European Fusion Development Agreement (EFDA) provides for collective use of the JET facility at Culham Science Centre, UK. The EFDA Work programme is organised in nine Task Forces. The Division is contributing by providing competence 7

8 in Task Force Heating (TFH) with specialists in Ion Cyclotron Resonance Heating, in Task Force Exhaust (TFE) and Task Force Fusion Technology (TFFT) with expertise in Plasma -Wall Interaction and in Task Force Diagnostics and Systems (TFD) in Spectroscopy Diagnostics and Diagnostics development. Additionally our Division is starting involvement in the work of Task Force MHD (TFM) with expertise in Resistive Wall Modes. The Division is co-ordinating several activities within the ITER-like Wall Project. The Division contributes to the JET Operator (UKAEA) tasks by supplying competence of Session Leader. A scientist seconded from the Division is contributing as Deputy Leader for Task Force Diagnostics with the opportunity to define and execute the EFDA-JET Work programme within the competence of Task Force D and as Manager for one of the Spectroscopy Enhancement projects. Other staff is seconded to the EFDA Close Support Unit (CSU) Culham as a responsible officer within the Programme Department, with special responsibilities for the Heating Task Force, and under a JET Operation Contract (JOC) with responsibilities for developing integrated tools and performing numerical analysis of JET plasma discharges, particularly with respect to ion cyclotron heating and in relation to the installation of the new ITER-like RF antenna. European task forces Fusion research conducted in the EU countries in certain strategic areas has recently been integrated into European Task Forces (TF). Two TFs have been formed so far: the Integrated Tokamak Modelling TF (ITM) and the Plasma-Wall Interaction TF (PWI). The Division is active in both these TFs. The Division leads a project for developing codes for heating, current drive and fast particle phenomena, one of seven projects within the ITM TF. It also co-ordinates several activities within the PWI TF related to the JET ITER-like Wall Project. ITER The Division of Fusion Plasma Physics contributes directly to the European ITER programme with the secondment of one scientist at the EFDA Close Support Unit (CSU) in Germany in the field of Physics Integration as liaison officer for ITER diagnostics. The Division is also involved in the EFDA Technology Programme within Physics Integration with specific tasks in the area Plasma-Wall Interactions and is involved in the development of RF heating through the ITER coordination committee for Fast Wave Heating and Current Drive. In addition, there are indirect contributions to ITER by participation in scientific programmes at JET that are dedicated to support ITER, including JET diagnostic upgrades. Collaborations with other EU fusion laboratories Experimental research in support of the ITER programme is carried out by the European associations at specialized national fusion facilities located throughout Europe. The Division is also involved in the research on active MHD control at Consorzio RFX, Italy, and at ASDEX Upgrade, Max-Planck Institut für Plasmaphysik, Germany. The Division is carrying out research on plasma wall interaction at the TEXTOR tokamak in Forschungszentrum Jülich (FZJ), Germany. 8

9 3. EXTRAP T2R experiment The Division of Fusion Plasma Physics is host for one of the European specialized fusion experiments, the EXTRAP-T2R reversed-field pinch device. One of the special features of the reversed-field pinch (RFP) configuration is that it is dependent on a close-fitting conducting wall for MHD stability. For a tokamak the stabilizing effects of a conducting wall are not vital, but they can improve stability for pressure driven modes and is used in the advanced tokamak configuration to enhance confinement performance. The front-end system of the previous EXTRAP T2 experiment (including the vacuum vessel and the conducting wall) was replaced during a rebuild in order to provide a new experiment that allowed reproducible operation with high quality plasma pulses that were substantially longer than the magnetic flux penetration time of the conducting wall. The goal was to provide a platform for studies of Resistive Wall Modes (RWMs), a special type of MHD instability that appears both in the RFP and the advanced tokamak configuration with a resistive wall. Figure 3.1. The EXTRAP T2R experiment at the Alfvén Laboratory EXTRAP T2R was upgraded during in collaboration with Consorzio RFX, Padova, Italy. A comprehensive system for active direct magnetic feedback control of non-axisymmetric MHD modes was developed and installed at EXTRAP T2R. The Active MHD Mode Control System installed has capabilities that at present are un- 9

10 paralleled in any tokamak fusion device in the world. In conjunction with other attractive features of the EXTRAP T2R experiment, the device, as it stands now, provides excellent capabilities for research on active control of MHD instabilities. The research program at the EXTRAP T2R device includes the following areas: MHD instability control. Non-linear MHD dynamics. Plasma turbulence. Plasma-wall interaction. Parameters of the EXTRAP T2R device are listed in Table 3.1: Table 3.1. EXTRAP T2R parameters Parameter Notation Value Unit Major radius R 1.24 m Minor radius a m Wall diffusion time τ v 6.3 ms Plasma pulse length τ d <100 ms Plasma current (typical) I p 100 ka Plasma electron temperature (typical) T e 300 ev Plasma electron density (typical) n e 1x10 19 m -3 Plasma diagnostics systems include Magnetic coil arrays for study of MHD activity. Electric and magnetic probe array for turbulence studies. Collector probes for plasma wall interaction studies. VUV and visible spectroscopy. Thomson scattering. Interferometer. Neutral particle time-of-flight diagnostic. Bolometer array. SXR camera. 10

11 3.1. Active MHD mode control system The active MHD mode control system installed on EXTRAP T2R has excellent capabilities. It is based on extensive arrays of active coils and sensors distributed over the toroidal surface of the resistive wall, shown in Fig Sensor flux loop inside shell (blue): norm. radius: r c /a=1.08 coil span: 90 o poloidal, 5.6 o toroidal Figure 3.2. Two-dimensional arrays of sensor flux loops and active saddle coils installed at EXTRAP T2R. The main features of the system are listed below: Active saddle coil outside shell (red): norm. radius: r c /a=1.3 coil span: 90 o poloidal, 11.2 o toroidal Sensor array. A two-dimensional array of magnetic sensors at 4 poloidal and 64 toroidal positions, measuring all three magnetic field components is available - a total of 4x64x3=768 magnetic sensors. Typically, radial field sensors at 4x32=128 positions are used. The radial field sensor is a one-turn radial flux loop spanning 90 in poloidal and 5.6 in toroidal direction respectively, measuring the radial magnetic flux at a radial position just inside the resistive wall. Active coil array. A two-dimensional array of active coils at 4 poloidal and 32 toroidal positions is installed outside the resistive wall - a total of 128 coils, providing a full cover of the wall. Each active coil is a 40-turn saddle coil that spans 90 in poloidal and 11.2 in toroidal direction, respectively. Saddle coils and sensor flux loops are pair-connected at each toroidal position to form 64 independent m=1 coils and sensors. Power amplifiers. A total of 32 power amplifier units with 2 channels each are installed, providing 64 amplifier channels in total. Professional audio amplifiers are used with output power of Watt and bandwidth 1 Hz to 25 khz. Amplifier output currents are up to 20 A, providing 800 At in the power coils and a maximum radial magnetic field at the coil centre of about 3 mt. 11

12 Integrated digital controller module. An integrated digital controller module, developed by Consorzio RFX, is presently used at EXTRAP T2R. The system is contained in a VME bus crate and includes 1) ADCs for analog input of 64 magnetic sensor signals and 64 coil current signals, 2) Board with PPC CPU, 500 MHz and 512 MB RAM, 3) DACs for analog output of 64 amplifier control voltages. Controller algorithms are implemented in software. Several control schemes are used at present, e. g.: o Intelligent shell : 64 independent sensor-coil systems with PIDcontrollers for active suppression of the total m=1 radial flux at the wall. o Mode control : Spatial FFT of 64 sensor signals and independent P- controllers with complex gains for active control of both in-phase and quadrature components for 32 individual m=1 Fourier modes. o "Open loop" mode for various studies involving application of preprogrammed stationary or rotating external controlled fields. Analog feedback controllers. An optional set of analog PID-type feedback controllers is available for use with the intelligent shell feedback control scheme EXTRAP T2R enhancements during 2006 A main enhancement to the EXTRAP T2R experiment during 2006 was the upgrade of the vertical field (VF) power supply and the VF feedback controller. The main features of the VF system upgrade were New large capacity parallel connected IGBT solid-state current switching elements. New controller allowing full PID action and separate vertical field (VF) and horizontal field (HF) control. The VF system is required for the equilibrium control in discharges extended over the shell time. Previously the maximum VF pulse duration was limiting the plasma discharge length to around 60 ms. The VF system upgrade has allowed the increase of the plasma pulse length with active MHD mode feedback control from around 60 ms in 2005 up to nearly 100 ms. 12

13 4. Summary of research results 4.1. Resistive wall mode control P. Brunsell, J. R. Drake, D. Yadikin (PhD student), E.Olofsson (MSc student) In collaboration with: E. Rachlew, A. Hedqvist, M. Kuldkepp (PhD student), S. Menmuir (PhD student), Dept. of Physics, SCI, KTH R. Paccagnella, G. Manduchi, A. Luchetta, Consorzio RFX, Padova, Italy Y. Liu, Radio & Space Science, CTH Performance improvement of high confinement tokamaks over currently achieved parameters will require active control of MHD instabilities. Active stabilisation is a relatively new field and the integration into scenarios with overall favourable properties in tokamaks is still to be demonstrated. Various control schemes have been proposed and tested, mainly in reversed-field pinch devices, and in particular experiments in the EXTRAP T2R device on resistive wall mode (RWM) feedback control have provided important initial results. The obtained experimental results are in basic agreement with the present theory models. These studies are of interest for the broad fusion community since they promote the understanding of basic aspects of RWM active control physics. The main points motivating the research program on active MHD instability control are, in summary: The development of active control of RWMs has high priority in the European fusion programme. If feasible, it will be a very attractive scenario for ITER, allowing a larger operating parameter regime and increased fusion power. The MHD mode control system installed on EXTRAP T2R has capabilities that at present are un-paralleled in any tokamak fusion device. In conjunction with other attractive features of EXTRAP T2R, the experiment provides unique capabilities for research in this field. The research project is carried out in collaboration with groups in Sweden (CTH) and in EU (Consorzio RFX, Italy), providing expertise in areas such as RWM modeling and real time systems engineering. Initial results of active control experiments in EXTRAP T2R have been very encouraging, providing 1. First experimental demonstration of successful active feedback control of multiple modes [P. R. Brunsell, et al. Phys. Rev. Lett. 93, (2004)]. 2. First demonstration of suppression of the full spectrum of unstable RWMs in the RFP using intelligent shell feedback [P. R. Brunsell, et al., Plasma Phys. Control. Fusion 47, B25 (2005)]. 13

14 3. Experimental verification of the linear cylindrical MHD model by measurement of both unstable mode growth rates and stable mode damping rates [D. Gregoratto, et al., Phys. Plasmas 12, (2005)] Intelligent shell feedback Starting from May 2005 experiments using the full array with 4x32=128 coils have been carried out. The initial results have been presented already at an invited talk at the EPS Plasma Physics Conference 2005, and the work has been continued during 2006, with emphasis on control system optimization using full PID controllers. The upgrade of the vertical field power supply carried during out during 2006 has allowed for longer plasma pulses and experiment is now routinely operated with active suppression of the full unstable RWM spectrum (of about modes) for more than 10 wall times [P. R. Brunsell, Nucl. Fusion 46, 904 (2006)]. The studies have focused on Feedback gain required to achieve full stabilization of the modes in the intelligent shell operation. The cylindrical linear MHD model is useful to describe RWMs in the RFP. In this model, unstable modes are individual Fourier harmonics described by poloidal and toroidal mode numbers (m,n). The linearly unstable RWM spectrum for a typical equilibrium in EXTRAP T2R consists of m=1 modes with -11<n<-2 and 1<n<6. The product of the mode growth rate γ n and the mode wall time τ n for the m=1 modes are shown in Fig γ n τ n Figure Cylindrical MHD model calculation of m=1 RWM growth rates. The figure shows the product of the growth rate and mode wall time. There is a range of about 16 unstable modes with positive γ n τ n. The highest growth rates are obtained for m=1 modes with n=-11,-10, corresponding to near on-axis resonance with the equilibrium field. Spontaneous growth of a range of RWMs is observed in EXTRAP T2R. Measurements of RWM growth rates are in good qualitative agreement with linear MHD stability calculations. As a result the active suppression of RWM, a three-fold extension of the pulse length is observed, Fig Studies of the feedback gain required to achieve full stabilization of the modes in the intelligent shell configuration have been performed. When the feedback gains are set equal for all coils, the operation is equivalent to mode control feedback, with equal gain on all modes. A preliminary simplified modeling neglecting the dynamics of the external circuits gives the minimum required feedback loop gain G for stabilization of a single mode as G>γ n τ n. The experiments have demonstrated that for the EXTRAP T2R feedback system there is a smooth improvement of mode suppression with increasing gain and 14

15 there is a range of gain values for which adequate suppression can be achieved with proportional gain. Introduction of integral gain reduces the required proportional gain, and derivative gain is useful to suppress the tendency for oscillations at high proportional gains. The effect of different feedback gains are shown in Fig A proportional feedback gain of G P =2 is sufficient for stabilization of the unstable RWMs with highest growth rates n=-11, -10,-8, 5, 6, in agreement with the modeling. However, in order to fully suppress some of the marginally unstable modes (e. g. n=2) that have high amplitude as a result of amplification of an external field error, a higher feedback gain (G P =10) is required. Figure From top: Plasma current, rms m=1 mode amplitude, mode amplitude for a centrally resonant tearing mode m=1, n=-13, mode amplitudes for dominant non-resonant RWMs n= -11,-10,-8,+2,+5,+6. Three discharges are shown: Without feedback control (black), feedback control with moderate gain G=2 (blue), feedback control with high gain G=10 (red). Discharge length is prolonged about three times with high gain feedback. A smooth improvement of mode suppression with increasing gain is observed, leading to successively longer discharges. Suppression of the full spectrum of unstable RWMs is obtained for the full discharge duration of about 10 wall times Linear coupling of side band modes A general feature of any MHD mode control system using a discrete array of coils is the linear coupling of side band modes. Coupling of unstable modes by the active coils will generally prevent simultaneous stabilization of all modes in a coupled set with the intelligent shell control scheme. The study investigates the Simultaneous stabilization of targeted modes and coupled "side-band" modes with the intelligent shell scheme. 15

16 The studies investigate the stabilization of targeted modes and coupled "side-band" modes. Different layouts of coil arrays have been obtained by using a part of the full coil array on EXTRAP T2R, typically consisting of half the number of coils (64 coils). [D. Yadikin, et al. Plasma Phys. Control Fusion 48, 1 (2006)]. Experiments are carried out using arrays of active coils resulting in different combinations of coupled modes, for example 4x16 coils or 2x32 coils. Modes with m=+1 and different toroidal mode number (-16 n 15) are considered. With a 2x32 system (instead of 4x16) different pairs of modes are coupled by the coils. With intelligent shell feedback, the dominant mode in the set decreases its amplitude while the other coupled mode in the set can grow. In the case of 4x16 coils, the mode n-spacing of the side-band is given by the number of toroidal positions (16) for the active coils. An example of a set of coupled m=1 modes is n= -10, +6, shown in Fig Amplitudes [mt] a) n=-10 Without FB b) n=6 Without FB With FB With FB Figure Example of toroidal side band effect with 4x16 coil layout giving coupling of the n=-10 and the n=+6 modes. The n=-10 mode is dominant so the side band effect leads to partial suppression of the n=-10 mode and faster growth of the n=+6 mode time [ms] In the case of 2x32 coils, the dominant coupling is between m=+1 and m=-1 modes. Considering only m=+1 modes, this corresponds to a coupling of modes with opposite sign of the toroidal mode number n. One example is the set of coupled m=1 modes n=-6,+6, shown in Fig Amplitudes [mt] a) n=6 With FB b) n=-6 With FB Without FB Without FB Figure Example of poloidal side band effect with 2x32 coil layout that couples the n=-6 and n=+6 modes. The n=-6 mode is dominant, and the side band coupling results in a partial suppression of the n=-6 and a faster growth of the n=+6 mode time [ms] 16

17 Fake rotating shell feedback The concept called fake rotating shell has been tested on EXTRAP T2R. This work is a First preliminary study of qualitative features of the fake rotating shell feedback scheme. The full sensor array of the radial field has 64 toroidal positions. By operating the control system in intelligent shell feedback mode with toroidally displaced sensors (instead of the using sensors at toroidal angles coinciding with the active coils) a phase shift is produced between sensor field and the control field [D. Yadikin, et al., Phys. Plasmas, 13, (2006)]. For low-n modes, the fake rotating shell feedback results in suppression of the modes as well as slow toroidal rotation. For high-n modes with n >16, the present preliminary experimental setup does not fulfill the condition that the sensor displacement should be small compared to the mode wavelength. This results in destabilization of the high-n modes at present limiting the effectiveness of this scheme Comparison of different magnetic field sensors Using the individual mode control configuration, the feedback gain can be set for selected harmonics, with both real and complex values. Using the mode control concept it has been possible to carry out the first feedback experiments using a sensor array measuring the tangential field component at the wall. The goal is this study is a Comparison of mode control feedback using toroidal and radial field sensors Figure Comparison of mode amplitudes (left) and phases (right) of n=-11 mode for three cases: Without feedback (black), with feedback using radial field sensors (blue), with feedback using toroidal field sensors (red). Radial field in top panels and toroidal field in bottom panels. 17

18 The mode suppression obtained with feedback using toroidal field sensors have been compared with that using radial field sensors, Fig The best suppression is obtained at the predicted complex gain phase (around π/2). Mode rotation is induced at other complex gain phases Mode control with complex feedback gain Using the mode control scheme with complex gain both the in-phase and the quadrature component of the mode is controlled. An arbitrary phase shift between the sensor field and the control field can be applied in order to affect the dynamics of the mode. The aim of this study is to investigate Simultaneous suppression and rotation of modes using complex feedback gains. Results of feedback with complex feedback gains are shown in Fig Experiments using varying magnitudes of the feedback gain and a fixed phase difference between the sensor field and active coil field show faster mode phase variation with increasing feedback gain. Mode amplitude Mode phase Total Coil vacuum Plasma Figure Mode control feedback of n=-11 using complex gains with different magnitude but fixed gain phase (π/6). Left: mode amplitude, Right: mode phase. From top: total field, control coil vacuum field, plasma field (total - coil). Cases shown are without feedback (black), with feedback using G=0.65 (red), G=1.3 (blue), G=2.0 (magenta). 18

19 4.2. Non-linear MHD dynamics and plasma turbulence J. R. Drake, M.Cecconello, H. Bergsåker In collaboration with: N. Vianello, V. Antoni, E. Spada, M. Spolaore, G. Serianni, R. Cavanazza, Consorzio RFX, Padova, Italy E. Rachlew, M. Kuldkepp (PhD student), S. Menmuir (PhD student), Dept. of Physics, SCI, KTH Tearing mode dynamics Tearing modes (TMs) form islands localized on rational magnetic flux surfaces that experience viscous forces from the plasma fluid such that the islands tend to follow the plasma flow. There is electromagnetic torques on the islands when the TMs interact with resonant magnetic fields created by induced wall currents or external field errors. Experiments have been carried out on EXTRAP T2R to study Tearing mode dynamics from measurements of mode velocity with magnetic coil arrays and spectroscopic measurements of plasma flow. TMs remain at a constant, non-linearly saturated level, spontaneously rotating at high velocities for a large fraction of the discharge duration. They behave qualitatively as if the shell were ideally conducting. This feature of the EXTRAP T2R device gives a clear separation of TM and RWM temporal behavior. In standard discharges without feedback, the TM rotation slows down prior to the discharge end, and the modes lock to the wall. One of the characteristics of plasma discharges operated with active RWM feedback stabilization is the prolongation of both the plasma rotation and the rotation of the core resonant TMs [M. Cecconello, Plasma Phys. Control Fusion 48, 1311 (2006)]. Figure Toroidal rotation of the core resonant TM n=-12 (solid line), (assuming zero poloidal mode rotation near the core), and the OVI ionization stage (solid circle) for two cases. Top: without feedback control of RWMs, Bottom: with feedback control There is a good correlation between toroidal mode rotation and plasma flow in the plasma core, as shown in Fig The poloidal mode rotation and the poloidal plasma flow at a larger plasma radius is also in agreement, Fig

20 Figure Poloidal rotation of the edge resonant TM n=-25 (solid line), (assuming parabolic radial profile of the toroidal angular velocity), and the OVI ionization stage (solid circle) for two cases. Top: without feedback control of RWMs, Bottom: with feedback control Plasma turbulence, flow and transport Plasma turbulence represents an outstanding issue for magnetically confined plasmas. Turbulence has been recognized at the cause of anomalous transport. The discovery of improved confinement regimes associated with turbulence reduction and sheared plasma flows has motivated an Investigation of the relationship between turbulence and the mean flow using probe measurements with high spatial and temporal resolution. The self-regulation model for turbulence can be described as follows: Electrostatic turbulence drives most of the mean flow through the velocity coupling term in the Reynolds stress tensor. ExB sheared flow is the result of the counteracting action of turbulence driving and viscous damping, where viscosity is anomalous occurring through electrostatic turbulence. The ExB flow shear is effective in reducing the coupling between density and velocity fluctuations, and it may modify the turbulence anisotropy and thereby the Reynolds stress In order to measure all terms in the momentum balance in EXTRAP T2R edge plasma, a probe array that combines electrostatic and magnetic measurements is used [N. Vianello, et al., Phys. Rev. Lett., 94, (2005)]. The probe array, shown in Fig , consists of a boron nitride (BN) case with 17 molybdenum pins and 2 three-axial magnetic probes. The five pins on the top of the BN case are used as a five-pin triple balanced probe, whereas the remaining ones measure floating potential. The 2 three-axial magnetic probes, 13 mm toroidally spaced, measure the time derivative of the three components of the magnetic field. Figure Probe array used in edge turbulence measurements in EXTRAP T2R. 20

21 As is commonly done in tokamaks and stellarators, fluctuating perpendicular velocities have been approximated by ExB drift velocity fluctuations. The electric field fluctuations have been derived from the gradient of the floating potential fluctuations. The measurements of the toroidal and radial mean velocity, their radial derivatives, and the fluctuations of velocity and magnetic field allows the main terms entering in the plasma momentum balance to be estimated, among which are the complete Reynolds stress and the triple correlation product. The momentum balance equation describes the momentum drive induced by fluctuations which is counteracted by the viscous damping of rotation. Using the measurement of the radial derivative of the Reynolds stress and the second derivative of the ExB velocity, an estimate of the perpendicular kinematic viscosity has been obtained from the momentum balance [N. Vianello, Plasma Phys. Control. Fusion 48 S193 (2006)]. The kinematic viscosity has the same dimension as a diffusion coefficient. In Fig , radial profiles of the experimental values (D exp ) are compared with the particle diffusion coefficient due to electrostatic fluctuations (D es ), which is also obtained from the probe measurements. Figure Radial profile diffusion coefficient (D exp ) and particle diffusion coefficient from electrostatic particle flux (D es ) The results suggest that electrostatic fluctuations play both the roles of flow drive through electrostatic Reynolds stress and flow damping through anomalous viscosity. A possible explanation for the different roles of electrostatic fluctuations could be that different instabilities lie behind the two processes. An analysis of the of the frequency spectrum for the in the momentum generation has been done. As is shown in Fig , the spectral components of the electrostatic Reynolds (ERS) stress peak at f=20-30 khz, and has been suggested to be related to m=0 tearing mode activity. The electrostatic particle flux exhibits a strong peak at f= khz, a frequency range which is far from the MHD activity. 21

22 Figure Frequency decomposition of the electrostatic Reynolds stress (full curve) and the electrostatic particle flux (dashed line). The energy transfer process and the role of ERS in generating mean kinetic energy has been studied by estimating terms in the mean kinetic energy balance equation. Of special interest is the mean energy production term P representing the energy exchanged from turbulent to mean flows. Positive P indicates that energy is transferred from fluctuations to mean flow, causing an increase in the mean kinetic energy content associated with plasma flow. The term is positive in the region of high velocity shear, Fig Comparing the contributions from electrostatic and magnetic fluctuations, it is observed that the electrostatic part dominates in the plasma region (inside the limiter radius). Figure Top: Radial profile of mean energy production term. Bottom: Radial profile of electrostatic and magnetic components of the production term. The limiter radius is indicated with a vertical line. 22

23 4.3. Plasma-wall interactions M. Rubel, P. Sundelin (PhD student), H. Bergsåker In collaboration with B. Emmoth, IMIT, KTH Plasma-wall interactions (PWI) comprise all processes involved in the exchange of mass and energy between the plasma and the surrounding wall. Two inter-related aspects of fusion reactor operation - economy and safety - are the driving forces for studies of PWI. The major issues to be tackled are: (i) lifetime of plasma-facing materials (PFM) and components (PFC), (ii) accumulation of hydrogen isotopes in PFC, i.e. tritium inventory; (iii) carbon and metal (Be, W) dust formation. PWI is one of the primary areas where integration of the Physics and Technology programmes is being achieved. The work at KTH in the field of PWI and fusionrelated material physics has been fully integrated with the international fusion programme: (i) EU Fusion Programme, (ii) International Tokamak Fusion Activity (ITPA), (iii) Implementing Agreements of International Energy Agency (IEA). It is demonstrated by the participation in: European Task Force on Plasma Wall Interactions (EU-TF-PWI). EFDA Technology Programme. EFDA-JET Work Programme: Task Forces E (Divertor Physics), FT (Fusion Technology), D (Diagnostics) and JET Enhancements (Phase 1 and Phase 2) including the ambitious ITER-Like Wall (ILW) Project, i.e. full metal wall at JET. ITPA and IEA activities. The research programme is concentrated on: Material erosion, migration and re-deposition. Fuel retention studies and fuel removal techniques. Characterization of plasma-facing materials and components including testing of high-z metals. Development and characterization of wall materials for ILW at JET. Development of diagnostic tools for PWI studies. Experimental work is carried out at home laboratory, JET and TEXTOR. EFDA-JET activity Studies of fuel retention and material migration On the last operation day with the MkII-SRP divertor 13 C-labelled methane was introduced to JET from the outer divertor. The aim was to study the carbon transport in order to better understand the mechanism responsible for the formation of thick carbon deposits in the shadowed region (i.e. area with no direct plasma line-of-sight) of the inner divertor. The analysis of the C-13 distribution and quantity on various plasma facing components (PFC: limiter and divertor tiles) were completed in The transport to the inner divertor was detected by ion beam analysis performed on the tiles retrieved from the torus. However, no marker was found in the shadowed region of the inner divertor. This indicates that the carbon transport to that region is a multi-step process. 23

24 Fuel retention studies have been continued for the tiles retrieved from the MkII-SRP divertor. The greatest fuel inventory has been measured in the shadowed region of the inner divertor Tile 4. The lack of beryllium in that region shows that the fuel inventory is associated with the co-deposition of carbon species transported over long distances. The results have been compared with those obtained for tiles from the MkII-A and Mk-IIGB divertors. There is a concern that the fuel inventory in fusion devices may also be associated with in-depth migration of hydrogen isotopes into the bulk of plasma-facing components, especially in carbon-based materials, especially in carbon fibre composites (CFC). JET has been operated with CFC components. A number of divertor tiles have been selected for analysis which is planned to be carried out in Development of Be coatings for the ITER-Like Wall Project The primary choice for ITER materials is a full beryllium (Be) main wall with CFC at the strike points and tungsten (W) at the divertor baffles and dome. Neither the reference material combination (Be Wall/CFC+W divertor) nor its backup (Be Wall/W divertor) have ever been tested in a tokamak. The lessons we can learn from such an experiment in JET may be critical for reducing the risk of material related problems in ITER. Therefore, the ITER-like Wall (ILW) Project has been launched at JET. The main aim of the Project is to operate a tokamak with a full metal wall: tungsten (W) divertor and beryllium (Be) components in the main chamber. This involves the use of: Uniform coating of the inner wall and upper dump plate cladding (Inconel) with a protective beryllium layer. Bulk Be limiters in the main chamber Outer Poloidal Limiters (OPL) and Inner Wall Guard Limiters (IWGL). In order to assess the erosion rate of Be from the limiters, it is necessary to develop methods of measuring erosion of a few microns during a campaign. A method of doing this is to distribute marker tiles as components of OPL and IWGL. The marker tile is a wall component of the same bulk material as the surrounding tiles (in this case Be), with a stripe of an easily detected heavy metal deposited on it as a thin interlayer, and with a few micron layer of the bulk material (i.e. Be) on top of that. It is shown schematically in Fig If the outer layer is eroded at the same rate as the bulk, then the erosion rate can be determined by detecting the distance of the interlayer from the final surface (for erosion of less than the film thickness). Good quality Be coatings are required for the Project. M. Rubel has been responsible for the overall coordination of the work carried out by several EURATOM Associations: FZJ- Germany, UKAEA United Kingdom, MEdC Romania, TEKES Finland, VR Sweden. 24

25 Be layer (7-9 μm) High-Z metal interlayer (2-3 μm of Ni) Beryllium Tile (3 cm) Figure Scheme of a marker tile structure. The ultimate goal of the R&D was to obtain and test coatings for Inconel and marker tiles in order to recommend the deposition procedure for a large-scale production of tiles for the ILW Project. The essential steps of the task were following: 1. Definition of the coating thickness and properties (purity) on Inconel. 2. Definition of the marker tile structure and layer properties (thickness, purity adhesion of the Be coating and metal interlayer). 3. Selection and optimization of proper deposition processes. 4. Production of test coupons on Inconel and Be tiles. 5. Pre-characterization of the products. 6. High-heat flux (HHF) testing to determine the performance (durability) of coatings under power loads. 7. Characterization of the coatings after the HHF test. 8. Recommendations for the large-scale production of tiles. Coatings on Inconel The layers were produced at the Nuclear Fuel Plant in Romania and the pulling test (for the adhesion determination) was performed. Morphology was characterized in order recognize the coating structure, thickness, uniformity and purity, especially the content of carbon and oxygen. The study was done by means of scanning electron microscopy (SEM), atomic force microscopy (AFM), energy dispersive X-ray spectroscopy (EDX), wavelength dispersive X-ray spectroscopy (WDS), glow discharge optical emission spectroscopy (GDOES) and ion beam analysis (IBA): Rutherford backscattering spectroscopy (RBS) with a proton and a 4 He beam and nuclear reaction analysis (NRA) with a 3 He beam. The original layer structure is shown in Fig a, whereas the adjacent image shows the layer after the high heat flux (HHF) testing at 18 MJ m -2. No changes of the surface properties have been observed at energy loads expected during the regular JET operation. a b 5 μm 5 μm Fig Be coating on Inconel: (a) original surface; (b) after HHF test (18 MJ/m 2 ) 25

26 Marker coatings on solid Be blocks Before the production of test coupons, the optimization of the deposition process for the interlayer and Be coating. The tests were done for Re, W, Cr, Ni-Re-Cr alloy and Ni on steel and silicon. The coatings were pre-characterized by means of SEM, EDS, AFM, Auger electron spectroscopy (AES), radiography and X-ray diffraction (XRD). SEM observations show that the Be layer is composed of tiny platelets (100 nm) of Be, see images in Fig , much smaller than obtained on Inconel by evaporation (see Fig ). In general, this indicates that the layer produced by thermionic vacuum arc method is more compact than the evaporated coating. However, there are some small holes (8-15 μm) in the layer and the Be platelets on edges of these holes are bigger (1 μm) than in the rest of the coatings. It has been difficult to assess whether the holes propagate through the entire coating thickness. Their origin is still to be understood. One possibility is that the holes were formed as a consequence of gas release (outgassing) from the material being maintained at the temperature of o C during the coating process. Figure Surface features of Be layer obtained by Thermionic Vacuum Arc Deposition and fuel inventory in C and Be MkI divertors (JET Fusion Technology Project) All plasma-facing components (PFC) in ITER will be castellated, i.e. composed of small blocks separated by narrow grooves (less than 1 mm) in order to reduce thermally-induced stress. The contact of plasma with several different elements (including carbon and beryllium) on the first wall, as foreseen for ITER, will also lead 26

27 to the co-deposition of eroded material together with fuel species in the castellation. The main question is whether fuel retention in such places may significantly contribute or even be decisive for the overall tritium inventory in ITER where the total number of castellated grooves will exceed 1 million. This calls for detailed studies of castellated structures and gaps between PFC tiles from present-day tokamaks. Until now, large castellated structures have been used at JET (divertor and limiters) and recently in Tore Supra (pump limiter). This contribution provides an account of the detailed examination of several castellated beryllium tiles from the belt limiter (of which approx pieces were in total) exposed to the JET plasma for s. The major aim of the investigation carried out by means of ion beam analysis methods (NRA and enhanced proton scattering [EPS]) was to determine the fuel retention and material mixing on the tiles. Analyses have been performed on both sides of the castellated grooves, on plasma-facing and side surfaces of the tiles. Images in Fig show the limiters cleaved along the grooves of castellation. The essential results are summarised by the following: Deuterium retention in the grooves is associated with co-deposition of carbon. Decay length of deposition in the castellation is short: λ ~1.5 mm. The maximum deuterium content in the groove does not exceed 8x10 17 cm -2 ; this is measured at the distance of approximately 2 mm from the entrance to the gap. No deuterium (above the detection limit of NRA technique) is detected in bulk Be. Bridging of gaps by molten beryllium is observed, but gaps are not filled with Be. Fuel content on plasma-facing surfaces is fairly low (in the range 1-8 x cm -2 ) mainly associated with carbon co-deposition. On side surfaces of the tiles the formation of BeO layer is detected at a distance of 20 mm and more from the plasma-facing surface. These results confirm our previous predictions that fuel inventory may strongly be reduced in a machine with a reduced carbon content on the first wall. Fig Beryllium limiters from JET First mirror test at JET for ITER First Mirror Test (FMT) project was initiated at JET in year 2002 within the Project on Tritium Retention Studies (TRS). The origin of this programme is related to the fact that first mirrors will be plasma facing components of all optical diagnostic systems in ITER. Mirror surfaces will undergo modification caused by erosion and redeposition processes influencing reflectivity of mirrors and thus affecting the spectroscopy signals. The limited access to in-vessel components of ITER calls for testing the mirror materials in present day devices in order to gather information on 27

28 the material damage and degradation of mirror performance, i.e. reflectivity. Therefore, the ITER Team requested the FMT experiment to be carried out. The Swedish EURATOM Association (Responsible officer: Marek Rubel) in co-operation with UKAEA carried out the project aiming at the manufacturing, delivery and installation of mirror samples and their carriers in several locations inside the JET vacuum vessel. All items were produced (see 2004 Report) and installed in the torus in They were exposed to the plasma during the campaigns in During the 2007 shut-down the mirrors will be retrieved from the machine and analysed. A new set of mirrors (produced in 2006) will be placed in JET. TEXTOR Material mixing on tungsten limiters Tungsten is a candidate material for the ITER divertor. In present-day fusion experiments both bulk tungsten (TEXTOR) and evaporated and plasma-sprayed coatings on carbon substrates have been employed (TEXTOR, ASDEX-Upgrade, JT- 60U). The coatings are also foreseen for the divertor and shine-through protection plates of JET in the ITER-Like Wall Project. The examination was carried out for tungsten coatings on graphite poloidal limiters exposed at TEXTOR: a vacuum plasma sprayed W layer ( mm) with a sandwich-type Re-W interlayer applied at the graphite tile to act as a diffusion barrier. When exposed to power loads of up to 20MW/m2 some limiters were partly damaged: melting and exfoliation of the coating occurred. The main focus of the study was on material mixing, i.e. the structure and composition of mixed phases formed in different areas of the coating: (a) melted zone and adjacent region; (b) nondamaged part coated by co-deposited layers. Following essential results obtained with X-ray diffraction (XRD), metallography (cross-sections) and several microscopy and surface analysis techniques have been obtained: In the overheated region adjacent to the melted zone XRD study has proven the co-existence of several major phases: metallic W, carbides (WC and W 2 C) and boride (W 2 B). Significant re-crystallisation of the coating and mixing of the W-Re interlayer have occurred in the overheated region: the regular W-Re sandwich interlayer (typical for the non-damaged part) has been transformed into a brittle intermetallic σ phase containing % of W. Transverse cracks being present in the phase layer usually end at the W Re/W interface. In some cases the cracks propagate along this interface causing delamination. Carbon, boron and silicon are the major plasma impurities in co-deposited films on top of the limiter. The content of deuterium ranges from 5x1014 cm-2 in the overheated area to 1x1018 cm-2 in carbon-rich films in the deposition zone. W-C-B mixing on the side surface of the limiter (i.e. surface in the gap between the graphite limiter blocks) suggests significant local transport of tungsten eroded from the coating. Material mixing leading to the new phase formation was also studied on a castellated solid tungsten limiter. As reported in 2005, the surfaces of castellated grooves contained deposits consisting of tungsten and oxygen. The surface structure is shown in Fig The analysis recently performed with XRD has allowed the identification of tungsten oxide (WO 2 ). This is probably the first-ever observation of 28

29 W oxide formation on PFC from a tokamak. Thermodynamic analysis is carried out to clarify the possible pathways of the oxide formation. The work on material mixing was performed in co-operation with the Polish EURATOM Association (IPPLM). Figure High resolution SEM image of the deposit containing W-O compound identified by means of XRD analysis as WO Fuel removal from PFC by laser irradiation Development of methods for fuel removal from PFC is one of the highest priority tasks, as defined in the EFDA and ITPA work programmes. The study has conducted co-operation with IPPLM and FZJ. The optimised parameters for Nd-YAG laser for the treatment of PFC tiles from the TEXTOR tokamak were following: ~0.6 J in 3 ns laser pulse, repetition rate of 10 Hz, shots in one series, beam diameter at the target ~3-4 mm. Optical spectroscopy and ion collectors allowed for in-situ monitoring of the process, whereas several material analysis techniques were used for ex-situ studies of the irradiated targets. The emphasis was on the studies of: (i) PFC surfaces before and after the irradiation and (ii) dust particle generated by the laser light impact. The major results obtained by means of microscopy, energy dispersive X-ray spectroscopy and nuclear reaction analysis are summarised by the following points. The irradiation removed a 50 μm thick co-deposited layer. The topography of the irradiated surface (after the co-deposit removal) is similar to that observed for an original (as produced) graphite. Dust particles were collected on surfaces of special dust catchers and on metal surfaces near the irradiated target. Their size distribution is near-gaussian with a mean grain size in the μm range. The dust contains deuterium, carbon and silicon as the main components. The presence of Si is related to the regularly performed siliconisation of the TEXTOR wall. The presence of deuterium in the laser-generated dust indicates that not all fuel species are transferred under the irradiation to the gas phase. This also shows that photon cleaning of PFC may result in the re-distribution of fuel-containing dust to the surrounding surfaces. Appropriate measures should be developed to minimise this effect. 29