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Magnetic Diagnostic Application for COMPASS Tokamak WDS'07 Proceedings of Contributed Papers, Part II, 229–233, 2007. ISBN 978-80-7378-024-1 © MATFYZPRESS Magnetic Diagnostic Application for COMPASS Tokamak D. Naydenkova,1,2 O. Bilyk,2 J. Stockel2 1 Charles University, Faculty of Mathematics and Physics, V Holešovičkách 2, 180 00, Prague, Czech Republic. 2 Institute of Plasma Physics, Association EURATOM/IPP.CR, Prague, Czech Republic. e-mail: [email protected] Abstract The paper outlines the main parameters and characteristics of typical plasma shapes. The reasons, basic physical principles and mechanisms for controlling of a plasma position in vertical and horizontal directions by means of magnetic sensors in the COMPASS tokamak are described below. All main parameters for the position control equate here. The sketchy description of diagnostic sensors and poloidal coils are used for control in COMPASS is done by the authors too. The progress report of current work is present as a summary. Introduction The COMPASS-D tokamak had been constructed and has been used until 2001 in the Culham laboratory in the United Kingdom and is now on move to Prague to be re-installed in the Institute of Plasma Physics. Its main parameters, working regimes and its application for its period of work in the Culham laboratory are presented for example in [Colton et al., 1999 and Fielding et al., 2000]. Magnetic sensors in tokamaks provide information about main plasma parameters such as plasma current, magnetic field, loop voltage, total energy, etc. Furthermore, they are used for monitoring the MHD activity in the plasma column because of their high frequency and spatial resolution. They are used to measure plasma position and shape and to control them by means of the feedback system. Controls in the COMPASS tokamak are realised by means of set of poloidal coils, which are shown in Fig.1. There are four types of poloidal coils used for different purposes: to drive an inductive current, equilibrium position, shaping of the plasma column and fast feedback control. Figure 1. Assembly representation of poloidal coils of COMPASS tokamak [The COMPASS project, 1983]. Plasma shape Plasma shape is characterised by two main parameters: elongation κ and triangularity δ . Elongation is defined as the ratio of the plasma half-height f to the plasma half-width a : 229 NAYDENKOVA ET AL.: MAGNETIC DIAGNOSTIC APPLICATION FOR COMPASS κ = f a Triangularity is the level of plasma resemblance with the D plasma shape (see Figure 2.): (c + d )/ 2 δ = a The main axis is drawn through the half-width a . Here, d is the displacement of the highest point of the separatrix, c - is the distance of the X-point to the main axis. Different combinations of poloidal coils are used for generating different types of plasma shape. The simplest is the circular one, which is always formed in the beginning of the discharges before applying any shaping field. The most common configuration is with a divertor (see Figure 2), where the separatrix separates well the confined plasma from the walls. This regime permits reaching better plasma purity and even confinement due to presence of strong magnetic shear in the vicinity of the X- point. It helps to protect the inner part of plasma from impurities, prevents plasma particles outflowing to the outside part. Although this process does not prevail over a plasma drift to the wall, it improves plasma confinement and its purity. Figure 2. The poloidal cross-section of COMPASS with divertor showing the magnetic configuration. [Bilykova et al, 2006]. Plasma position control The Plasma Position Control System is used to establish the equilibrium position of the plasma column inside the vessel in horizontal and vertical directions. In the COMPASS-D tokamak the control is realised by means of two control systems with different system response: “Slow” pre- programmed control, and “Fast” control (~5 kHz) which applies strong external forces driven by currents in the poloidal coils, controlled in real-time using magnetic measurements of the plasma position. Horizontal position control The toroidal plasma column with a current has tendency to increase its major radius R0 (to expand) because of kinetic and magnetic pressure. This increase has to be balanced by applying a r horizontal force FR : r r r FR ∝ j × Bz , r where Bz is vertical magnetic field, j toroidal plasma current density. The required equilibrium Bz is given by the expression [Wesson J, 2004]: 230 NAYDENKOVA ET AL.: MAGNETIC DIAGNOSTIC APPLICATION FOR COMPASS µ0I 8R0 1 Bz = − (ln + Λ − ) , 4πR0 a 2 l where Λ = β + i −1 pol 2 β pol – ratio of plasma and poloidal magnetic field pressures, li - ratio of square average poloidal magnetic field and square poloidal magnetic field at the edge, I - plasma current. Schema of sensors used for horizontal position control in COMPASS tokamak is shown in Fig. 3. Signals from two toroidal loops Loopin and Loopout located at the midplane are used for such control. To determine vertical magnetic field Bz the differential signal from the two loops is used. Pick-up coil is located inside the vacuum vessel at the midplane of the outer side of torus. The integral signal from the pick-up coil is proportional to the plasma current and horizontal displacement of the plasma column. Resulting signals are processed in the control system (either analogue or digital) and used as input for power supplies (PS). The output of PS is connected to poloidal coils E (slow control) and F (fast feedback control). The resulting control field is shown in Figure 3. F E / F MB Pick-up Coil Loopout Loopin F E / F Figure 3. Scheme of horizontal position control in the COMPASS tokamak [The COMPASS project [1983]. Vertical position control Elongated plasma is unstable to any displacement in the direction of the elongation. The criterion for stability against a vertical motion is [Wesson J, 2004]: 2 b + a ⎛ b + a ⎞ D = ⎜ ⎟ −1 < 0 b − a ⎝ b'+a' ⎠ a, b – minor and major semi-axes correspondingly; a’, b’ – semi-axes of perfectly conducting elliptical shell. If the surrounding shell has a finite conductivity, inertial instability appears with a growth rate −1 γ : γ = (Dτ R ) where τ R is resistive time penetration of the shell. 231 NAYDENKOVA ET AL.: MAGNETIC DIAGNOSTIC APPLICATION FOR COMPASS Figure 4. Schema of poloidal magnetic field lines. Plasma elongation is created by means of 4 poloidal coils shown schematically in Figure 4. If the plasma is displaced in a vertical direction, a perturbed vertical force due to the interaction between the poloidal field coil current and the plasma current appears. If this force increases the displacement of plasma, it leads to instability. To protect plasma from disruption, corresponding radial magnetic field have to be applied. Direction of this force depends on plasma displacement. If plasma moves upwards, the controlling magnetic field must point towards the low field side and vice versa. Pick-up Coils Toroidal Loops Poloidal Control Coils Figure 5. Scheme of vertical position control in the COMPASS tokamak [Vyas P, 1998]. 232 NAYDENKOVA ET AL.: MAGNETIC DIAGNOSTIC APPLICATION FOR COMPASS Schema of sensors used for vertical position control in the COMPASS tokamak is shown in Figure 5. For such control vertical force has to be applied: r r r Fz = j × Br Position of the toroidal loop sensors are shown in Figure 5. For determination of radial magnetic field Br, the integrated sum of signal differences from these toroidal loops is used (see figure 6). Pick- up coils are located inside the vacuum vessel symmetrically relative to the midplane as it is shown in Figure 5. The integrated sum of signal differences from the pick-up coils is proportional to the plasma current and horizontal displacement of the plasma column. Resulting signals are processed in the control system and used again as input for the power supplies (PS). The output of PS is connected to poloidal coils S (slow control) and F (fast feedback control). The resulting control field Br is shown in Figure 6. Toroidal loops B Ф = Si×Br = 2πRh×Br Figure 6. Measurement of radial magnetic field by means of toroidal loops. Here they are shown only two toroidal loops. Si – is the surface of the integration, R – is the distance from the main axis to the poloidal loop, h – is the distance between the toroidal loops. Progress report Nowadays, the work connected with the COMPASS transport to Prague is carried out. The building for hosting of the COMPASS tokamak is underway. It will be finished at the end of the year 2007. A majority of equipment has already been transported to Prague. The transportation of tokamak from Great Britain is planned in August 2007 and first shot is planed at the end of 2008. One of the main tasks for this moment is re-installation of the existing feedback system. It means tests of integrators, waveform generators etc. Gradual replacement of existing analogue feedback system to a digital one is envisaged. References. Colton A.L. et al, Error field penetration and plasma rotation in H mode and control of ELM-free regimes in COMPASS-D. Nuclear Fusion, vol.39, No 4, 1999. Fielding S.J. et al, Divertor Target Heat Load Reduction by Electrical Biasing, and Application to COMPASS-D. European Physical Society, Budapest, Hungry, June 12-15, 2000. The COMPASS project documents. Part II. UKAEA, Culham laboratory. December 1983. Bilykova O., et al, "COMPASS-D magnetic equilibria with LH and NBI current drive", Czechoslovak Journal of Physics 56: B24-B30 Suppl. B 2006. Wesson J., “Tokamaks”, International series of monographs on physics, Oxford, 2004, ISBN 0-19-850922-7. Vyas P., Vertical position control on COMPASS-D, Fusion Technology, 33(1998), 97-105.
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