The COMPASS tokamak Jan Stockel and the COMPASS

The COMPASS tokamak Jan Stockel, and the COMPASS team Institute of Plasma Physics, Prague Czech Republic MEPh. I, 7 April 2016 1

OUTLINE • Brief history of tokamak research in the Czech Republic (TM 1 – MH, CASTOR, GOLEM, COMPASS) • Introduction of the COMPASS tokamak (Hardware, additional heating, diagnostics, Data acquisition, Database, …) • Scientific mission of COMPASS • Typical discharge regimes Ø Ohmic regime ØH mode, ELMs • Additional plasma heating by Neutral beam Injection 2

History of fusion research in Czech Rep. 1963 -1974 Study of magnetized plasmas in linear facilities 1974 -1985 Tokamak TM-1 -MH (from Kurchatov Institute) 1985 -2005 CASTOR (reconstructed TM-1) - edge plasma studies by probes 2009 -till now GOLEM tokamak (reconstructed CASTOR) at Czech Technical University - for education – main feature – remote operation via Internet is possible 2005 Project of re-installation of the COMPASS tokamak approved by EURATOM 2006 -2009 2010 New building, power supplies, transport of COMASS from Culham lab to Prague (~12 MEuro) First plasma achieved 2011 -2012 Diagnostics development, discharge tuning, Neutral Beam Injection…. 2012 H mode achieved 2013 -2016 Full scientific operation 3

The COMPASS tokamak Small size, but ITER relevant geometry & magnetic configuration Typical cross section => D-shape Operational at IPP Prague since 2009 Major radius [m] 0. 56 Minor radius [m] 0. 2 Plasma current [k. A] < 350 Magnetic field [T] < 1. 15 (2. 1) Triangularity ~ 0. 4 Elongation ~ 1. 8 Pulse length [s] < 0. 5 Working gas Separatrix Core plasma , D 2, H 2 He Plasma cross section => circular, elongated, D-shape All orientation of Bt an Ip are possible Scrape off layer (SOL) X point Divertor 4

Mission of COMPASS Main goals • Plasma with Te ~ Ti • H-mode at these conditions • Get ELMs – possibly of Type 1 • Study edge plasma in detail COMPASS • Develop edge plasma diagnostics • Study ELMs mitigation • Perform scaling studies • Training To fulfil the scientific programme • Neutral Beam Injection are installed for ion heating • A new digital feedback system for plasma control 5

Tokamak vessel Vessel INCONEL Belt limiter Graphite tiles Inertial cooling only 64 Diagnostic ports Divertor Graphite tiles Inertial cooling only 6

Poloidal coils M winding – plasma breakdown and plasma current drive E winding – equilibrium position of the plasma column S winding – shaping of the plasma column F winding – fast control of plasma position (feedback system) 7

Circuit Waveform and timing Start of toroidal magnetic field At t = -1 sec Puff of working gas Start of NBI Start of flat-top End of NBI End of flat-top Start of shaping Breakdown Start of discharge End of discharge 8

Power supplies - schematically Of about 70 MW is required to drive current pulses of pre-defined shape and amplitude. But, only ~1 MW is available! For TF coils (~90 k. A), plasma breakdown, plasma current, equilibrium, shaping and additional heating systems Energy storage Tranformer 6 k. V => 600 V Thyristor rectifier (pre-programmable) Linkboard High current cables COILS Fast control of the plasma position and plasma density Grid Fast amplifiers (feedback controlled) Digital feedback control Sensors of plasma position & density The PS complex is manufactured and commissioned by a single Czech company 11

Flywheel generators - converters 1200 tons of reinforced concrete used to accumulate vibration during fly-wheel breaking • noise outside 56 d. B Start up ~ 40 min Recharge ~ 15 min Rotation speed 1700 - 1300/min Power 35 MW Frequency 85 Hz - 65 Hz Usable energy 45 MJ Total mass 52 tons electric drive 200 k. W 14 almost identical pre-programmable rectifiers are manufactured: 8 for TF coils and 6 for breakdown, magnetizing, equilibrium and shaping currents) The Power Supply complex is manufactured and commissioned by a single Czech company

Vacuum and wall conditioning v Standard pumping system is exploited • two turbo-molecular pumps (2 x 500 l/s) + a for-vacuum pump • Important – vacuum system has to be “oil-free” (special quite expensive vacuum pumps are required! • Basic pressure – 10 -8 mbar (10 -6 Pa) v Baking of the tokamak vessel to 150 o after opening to atmosphere (for few days) v Glow discharge (in He) in between every tokamak discharge – reproducibility v Boronization of the first wall is required to achieve the H mode - glow discharge in He with vapors containing boron (~ once per month) 13

Plasma diagnostics Magnetics (400 sensors) Loop voltage, plasma current, plasma position, EFIT reconstruction + arrays of Mirnov coils Microwave interferometer (4 mm) – line average density Spectroscopic diagnostics • Temporal evolution selected spectral lines - Da , impurities (C, He, . . ) • Several fast cameras (visible, infrared) • HR Thomson scattering core + edge – radial profiles of the electron temperature and density 52 spatial channels, temporal resolution 17 ms • 6 Arrays of AXUV-based fast bolometers • 3 arrays SXR detectors Probes • Reciprocating manipulator equipped with several probe heads • 39 Langmuir probes embedded in divertor tiles Neutral particle Analyzer – temporal evolution of fluxes of fast neutrals 0, 5 – 40 kev, ion temperature with temporal resolution of 50 ms Lithium beam diagnostics - temporal evolution radial profiles of plasma density at the edge of the plasma column Other diagnostic tools – ECE emission, several spectrometers (VUV, visible, near IR), HXR and neutron detectors … 14

Diagnostics - layout 13

Array of magnetic and SXR sensors F • Array of magnetic coils measuring poloidal magnetic field fluctuations, spaced toroidally • High frequency fluctuations: MC-A and MC-B and MC-C Low frequency fluctuations: IPR array • • courtesy Tomas Markovic Arrays of AXUV Bolometers and soft X-ray detectors 16

Data acquisition system COMPASS is equipped by ~ 1000 channels connected to Analog-Digital Convertors Sampling frequency 50 k. Hz - technologic channels (current and voltages in poloidal field coils) 2 MHz - majority of diagnostics 5 -10 MHz – special diagnostics (probes) 200 MHz – very special Thomson scattering, reflectometry In total, several Gigabyte of data is stored during a single discharge Data are organize in COMPASS Data. Base (CDB) for every discharge Express information is available on the COMPASS logbook in a few minutes after the discharge 115

COMPASS control room 16

Plasma performance at D-shape plasma cross section Magnetic coil system SND SNT Scrape-of layer Separatrix REALITY Strike point at outboard side X - point 17 Strike point at inboard side

Magnetics and plasma shape reconstruction 400 magnetic coils available! Some of them used for • Plasma position control • EFIT reconstruction of magnetic surfaces 18

Evolution of a typical Ohmic discharge Record of the shot #11 729 from the COMPASS logbook Plasma current ~ 180 k. A Shaping current ~ 5 k. A Line average density ~ 4 1019 m-3 Interferometer + Thomson scattering Loop voltage ~ 1, 5 V Intensity of Da spectral line

Global energy balance Ploss unknown underlying physics conduction, convection, …. Therefore, such power losses are characterized by the global energy confinement time t. E Global energy confinement time can be determined in the stationary phase of a discharge as

What is the H mode Plasma Pressure Radial profile of plasma pressure - schematically Normal tokamak operation - L mode • Radial profile of plasma pressure is parabolic with maximum in the center of the plasma column • Heat conductivity is anomalous elsewhere due to plasma turbulence k ~1 m^2/s Improved confinement – H mode • A narrow region with a steep pressure gradient is formed in the edge plasma (a phase transition) • Transport barrier, where the particle and heat transport is less turbulent (almost ”classical”) because of amplification of the radial electric field at the plasma edge Heat transport in the core plasma remains unchanged – it is still turbulent Plasma center Plasma edge The total kinetic energy in the plasma increases, and consequently, the global energy as well as the particle confinement improves!! Spontaneous phase transition – not fully understood yet

ELM free H-mode #5997 • Manifested by a sudden drop of Da emission • Plasma density increases – improved particle confinement • Occurs when plasma energy ~ 4 k. J • Loop voltage and consequently POH = Uloop * Iplasma increases because of improved particle confinement – accumulation of impurities • Increase of radiated power • ELM free discharges are usually terminated by disruption

Radial Profiles by HR Thomson scattering #6358 PNBI = 515 k. W Electron density Electron temperature Black symbols – L mode Blue symbols – H mode • Electron pressure ~ ne x Te • • Formation of pedestals on Te and ne are clearly visible Electron heating after LH transition Electron density increase as well • • courtesy Estera Stefanikova

Edge Localized Modes P l a s m a p r e s s u r e Transport barrier becomes unstable, if the pressure gradient is higher than a certain value => The pressure gradient drops down (almost up to the L mode level) => A significant fraction of plasma energy (~5 – 15 %) and particles flux is released to the wall in the time scale ~1 ms => The transport barrier (H mode) recovers These transient features (quasiperiodic) are called ELMs

ELMy H-mode #9329 Plasma current - ~ 300 k. A Line average density ~ 4 -5 x 1019 m-3 Loop voltage ~ 1 – 2 V Da emission, HXR emission

Typical Ohmic H-mode shots # 6317 ELM free phase – plasma density ramps up Less frequent ELM – density is stabilized More frequent ELMs – plasma density ramps down 26

Requirements for transition to H-mode COMPASS recipe o achieve the H mode • Divertor configuration with X point – elongation 1. 8 • Boronized vessel – the first wall is covered by a thin layer of Boron. An advanced technology is used: vapors of Carborane powder (C 2 B 10 H 12) are added to helium cleaning glow discharge for ~ 4 -6 hours (vessel at T = 150 o). Main advantage of Carborane- not flamable, not explosive, not poisoning!! • Deuterium as working gas Discharge conditions - found experimentally • Plasma current > ~ 180 200 k. A. • Corresponding ohmic power POH > 200 k. W • Line average plasma density > ~ 4 1019 m-3 • Additional plasma heating with the Neutral Beam Injection helps

Additional plasma heating with NBI Additional plasma heating with 2 Neutral Beam Injectors Beam energy (H, D) 40 ke. V Total power max. > 2 x 0. 3 MW Beam diameter < 5 cm Pulse length View from the top < 0. 3 s 1 I NB I 2 NB The beam is injected tangentially to plasma in the same direction as the plasma current => co-injection 5

Injector of the beam of neutral atoms schematically D 2

Neutralization of the ion beam Accelerated ion beam is neutralized in the molecular gas target in the neutralizer (filled by the working gas from the ion source). Beam particles do multiple collisions In a sufficiently "thick" gas target (given by n. H 2*Lneutr) charge exchange =>ionization=>charge exchange=> ionization=>. ) Ratio of fast neutrals and ions at the output of the neutralizer is just the ratio of the cross sections of the charge exchange and ionization in the molecular gas F = ex / ion Beam energy 40 ke. V (COMPASS) Theoretical efficiency of neutralization For Deuterium beam is around 80 %

NBI on COMPASS Weight ~ 2. 5 ton Cost ~ 1 MEuro

Extraction and focusing of ions Ions are extracted from the RF source by a system of four grids and accelerated to 40 ke. V The negative voltage on the grid G 3 repels electrons of the “secondary” plasma (since the fast beam ions ionize the neutral gas in front of G 4)

Extraction and focusation of ions from the ion source • • • Each grid has 887 holes – diameter 4 mm in fact – we have 887 beams Grids are cooled by water only at the edge Dominant cooling mechanism – radiation between shots. Otherwise, the shape of the grid is deformed (up to several tenth of mm) Surface of the grids must be conditioned step by step (increasing beam energy and ion current) Ion beam diameter 167 mm

NBI at tokamak area Vacuum tank Gas handling system Ion source Water cooling manifolds

2 x NBI on the COMPASS tokamak NBI-1 now out of operation NBI-2 Iplasma I plasma Co-injection: Neutral beams are injected into the plasma tangentially in the same direction as the plasma current The beam path in the plasma ~ 1 m Conter-injection: Reversal of plasma current and Btor Energy of the neutral beam 40 ke. V Extracted ion current 10 -12 Amps Efficiency of neutralization 80% Neutral beam power at the output of NBI PNBI = 40 ke. V x 10 Amps x 0. 8 = 320 k. W

Picture of the neutral beam The neutral beam shoots into the COMPASS vessel filled by deuterium gas Interaction of the beam With the port is evident