Reconstruction of plasma shape and plasma energy in Spherical Tokamak Globus-M.

1. Reconstruction of plasma shape and plasma energy in Spherical Tokamak Globus-M. S. Bender1, V.Gusev, A.Detch, Yu.Kostsov1, R. Levin, K. Lobanov1, N.Sakharov. A.F.Ioffe Physico-Technical Institute, St.Petersburg, Russia 1 D.V. Efremov Institute of Electrophysical Apparatus, St. Petersburg, Russia

2. Spherical tokamak Globus-M Tokamak Parameters: • Major radius R =0.36m • Minor radius a = 0.24m • Aspect ratio A > 1.5 • Achived Plasma Current Ip < 370kA • Toroidal magnetic fieldBT = 0.07 - 0.55Тл Achieved plasma parameters: • Plasma density <n> < 1.21020m-3 • Electron Temperature Те < 1000 эВ • Ion Temperature Тi < 600 эВ • Elongationk = 1.1 – 2.2 • Triangularity  = 0.1 – 0.45 • Safety factor q95  2.1

3. Magnetic diagnostic analyses

4. Magnetic loops quantity and coordinate selection requirements 1. Quantity of magnetic loops must be minimum 2. Magnetic diagnostic must work If 1-2 loops are damaged 3. Magnetic loops cannot cross the aperture of ports 4. The reason for selection of magnetic loops coordinate is minimization of magnetic flux extrapolation error to closed circuit. for measurements near ports and other diagnostics for poloidal field reconstruction before plasma discharge

5. Upgrade of magnetic diagnostic Globus-M magnetic diagnostics • 9 Rogowsky coils for measurements current in tokamak coils • 21 magnetic flux loops • 64 poloidal magnetic field detectors • Rogowsky coil positioned outside the vacuum vessel for vacuum vessel current measurement • Rogowsky coil positioned inside the vacuum vessel for plasma current measurement • 2 diamagnetic flux loops Magnetic flux loops In winter of 2004-2005 14 additional magnetic flux loops closed toroidally were installed inside the vacuum vessel

6. Globus - M model PF coils and magnetic flux loops In the model we use real dimensions, coordinates and the number of turns for each PF coils and magnetic flux loops.Besides the signals of the magnetic flux loops the EFIT input data include currents in the PF coils and the induced toroidal current in the vacuum vessel Limiter In the limiter magnetic configurations the plasma outmost closed magnetic surface is determined by the graphite limiter on the vessel central cylinder.

7. Globus - M model Vacuum vessel The vacuum vessel is an all-welded stainless steel construction with the characteristic wall thickness 2-3 mm except the outer ring of 14 mm thickness. Due to the small wall thickness the vessel electric resistance to the toroidal current is about 0.1 mOhm. For this reason the current flowing through the vessel in the toroidal direction does not exceed 40-50 kA in plasma current ramp-up phase and 15-20 kA during the plasma current plateau. The last values are small in comparison with the plasma current values. However, the spatial distribution of the vacuum vessel current is taken into account. For this purpose the vessel is extrapolated by the set of 21 rings corresponding to the number of loops. Current in each ring is determined according to the induced voltage measured by the closest loop.

8. Central Solenoid The central solenoid is extrapolated with two-layer coil wound by the condactor of 20×20 mm×mm cross section with a uniform current density distribution. In the experiments the central solenoid operates in current swing regime. The plasma breakdown is initiated in phase of positive current ramp down. The starting value of the current is 50-55 kA per turn. The plasma current plateau is sustained during the solenoid negative current half-wave, where current reaches the values of 40-45 kA at the end of plasma shot. In this phase the central solenoid stray magnetic field plays a significant role in the formation of plasma magnetic configuration despite the energizing of compensation coils for the stray field correction. Especially important the stray field asymmetry relatively the tokamak midplane caused by some inaccuracy in the solenoid manufacture and the tokamak assembly. The adequate model describing the errors in the solenoid constraction should be developed on the base of magnetic measurements.At present the central solenoid asymmetry is described by the coil vertical shift of 3.5 mm above the midplane. Globus - M model

9. Formation of detached magnetic configuration • A systematic observation of reconstructed plasma magnetic configurations revealed that both X-points are usually located inside the vacuum vessel volume during a most part of the plasma current plateau phase. At the same time the plasma is usually limited by the vessel wall and the outmost closed magnetic surface is determined by the plasma contact with the graphite limiter on the vacuum vessel inner cylinder near the tokamak midplane. The formation of a fully detached plasma was demonstrated in the experiments. • The transition from limiter to X-point configuration was accompanied by some increase of the plasma vertical elongation. Another evidence of the plasma detachment is a significant decrease in the intensity of impurities emission observed by collimated detectors in the midplane. Figure illustrates at least by a factor of 3-4 drop in the intensity of the OIII and CIII lines during a few milliseconds interval. • Figure shows also the variation of plasma thermal energy obtained from EFIT analysis. However, for further study of plasma energy balance the EFIT energy must be verified by kinetic measurements of the plasma temperature and density spatial distributions.

10. Monitoring of the plasma vertical position • Before the installation of 14 additional magnetic flux loop the EFIT analysis was based on the experimentally measured values of the plasma current, currents in PF coils and the plasma radial position. The reconstruction was performed for magnetic configurations symmetrical relatively the midplane. First measurements using new magnetic loops have demonstrated a tendency to the plasma vertical shift towards the vessel lower dome and a formation of a single null configuration as the amplitude central solenoid negative current increased. Most plasma shots were terminated by the internal reconnection event (IRE) accompanied by vertical disruption even at large safety factor values. =>

11. Monitoring of the plasma vertical position • The input signal for the plasma vertical position control was a radial magnetic flux measured by two loops positioned on the top and lower vessel domes. The plasma vertical position was monitored according to the preprogrammed radial magnetic flux waveform. Usually the radial flux value closed to zero was chosen for the steady state phase of the plasma discharge. However, the EFIT analysis revealed a strong, up to 10-15 cm, vertical displacement of the plasma geometrical center when the automatic system sustained the radial magnetic flux according to the preprogrammed zero value. This effect is caused by the superposition of the asymmetric relatively the midplane central solenoid stray field and the radial symmetric magnetic field produced by the horizontal field coils in the feedback control contour. The absolute value of the plasma vertical shift was larger for larger values of the plasma vertical elongation. Note, that IRE occurs in plasma with higher vertical elongation and therefore with higher value of the safety factor near the plasma boundary, but at larger vertical shift.

12. Monitoring of the plasma vertical position • The plasma column can be shifted back towards the center of the vacuum vessel by a correction of the preprogrammed radial magnetic flux wave form. The time evolution of the plasma current and plasma vertical position in the shots with different programs of radial magnetic flux. In most shots the correction of the plasma vertical position led to an increase of the plasma shot length.

13. Measurements of plasma total stored energy Elongated plasma cross-section (k>1.3) lets to reconstruct total plasma energy.

14. Conclusions • A successive upgrade of magnetic diagnostics have been performed in Globus-M experiments. A new diagnostic together with improve model of the tokamak magnets and the vacuum vessel made possible EFIT reconstruction equilibrium magnetic configurations of different types by using a small number of magnetic loops. The magnetic loops and the specially developed fast data acquisition system will be employed further for a real time digital control of the plasma shape. • Detached plasmas limited by a single null separatrix were achieved in tight plasma – vacuum vessel configuration. • The influence of the central solenoid asymmetric stray field on the plasma shape was studied. The application of additional radial magnetic field in the plasma current plateau phase led to the plasma stability enhancement and as a result to an increase of the plasma shot length. • This work was supported by Rosatom and RFBR grants (03-02-17659, 05-02-17773)