The W7-X Team. Theorist s view on Wendelstein 7-X (from its top; blue: coils, yellow: plasma surface)

Transcription

1 The W7-X Team Alonso [1], Andreeva [2], Baldzuhn [2], Beurskens [2], Beidler [2], Biedermann [2], Blackwell [17], Blanco [1], Bosch [2], Bozhenkov [2], Brakel [2], Burhenn [2], Buttenschön [2], Cappa [1], Czarnetzka [3], Dinklage [2], Endler [2], Estrada [1], Fornal [3], Fuchert [2], Geiger [2], Grulke [2], Hartmann [2], Harris [4], Hirsch [2], Hoefel [2], Jakubowski [2], Klinger [2], Klose [2], Knauer [2], Kocsis [5], König [2], Kornejew [2], Krämer-Flecken [6], Krawczyk [3], Krychowiak [2], Kubkowska [3], Kiazek [7], Langenberg [2], Laqua [2], Laqua [2], Lazerson [8], Maaßberg [2], Marsen [2], Marushchenko [2], Moncada [9,10], Moseev [2], Naujoks [2], Otte [2], Pablant [8], Pasch [2], Pisano [11], Rahbarnia [2], Riße [2], Rummel [2], Schmitz [12], Schröder [2], Stange [2], Stephey [12], Szepesi [5], Trimino-Mora [2], Thomsen [2], Traverso [13], Tsuchiya [14], Turkin [2], Velasco [1], Wauters [15], Werner [2], Wolf [2], Wurden [16], Zhang [2], et al. [1] CIEMAT, Madrid (Spain) [2] Max-Planck-Institut für Plasmaphysik, Garching, Greifswald (Germany) [3] IPPLM, Warsaw (Poland) [4] Oak-Ridge National Laboratory, Oak Ridge, TN (USA) [5] Wigner RCP, Budapest, (Hungary) [6] Forschungszentrum Jülich, Jülich (Germany) [7] Opole Univerisity, Opole (Poland) [8] Princeton Plasma Physics Laboratory, Princeton, NJ (USA) [9] CEA, Cadarache (France) [10] ThermaVIP Ltd., Cadarache (France) [11] University of Cagliari, Cagliari (Italy) [12] University of Wisconsin, Madison, WI (USA) [13] Auburn University, Auburn, AL (USA) [14] National Institute for Fusion Science, Toki (Japan) [15] ERM, Brussels (Belgium) [16] Los Alamos National Laboratory, Los Alamos, NM (USA) [17] Australian National University, Canberra (Australia) T Theorist s view on Wendelstein 7-X (from its top; blue: coils, yellow: plasma surface) Andreas DINKLAGE for the W7-X Team First Experiments on W7-X NIFS, Toki 30. May 2016 Page 1

2 Summary Optimized Stellarators: a potential path to a fusion power plant (FPP ) W7-X is to bring the HELIAS line to maturity (FPP) Qualify key technology Show good plasma confinement & proof optimization Demonstrate safe high-performance steady-state operation (heating, fuelling, exhaust, impurities, fast-ions, MHD, scenarios) (some selected) diagnostics, their requirements and applications: results from the first operation phase Diagnostics are very much as in tokamaks, accents on 3D, steady-state, ECRH possible contributions to ITER, DEMO W7-X is comprehensively equipped with diagnostics allowing to address first aspects of the high-level vision of W7-X Diagnostics for S-DEMO? Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 2

3 Diagnostics Developments for W7-X (not all!) Potential Applications for Burning Plasma Devices International School of Fusion Reactor Technology, Andreas Dinklage 1,2 for the W7-X Team 1 Max-Planck-Institut für Plasmaphysik, Greifswald, Germany 2 E.-M.-Arndt Universität Greifswald with gratitude to R. König, D. Hartmann and S. Bozhenkov This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the European Union s Horizon 2020 research and innovation programme under grant agreement number The views and opinions expressed herein do not necessarily reflect those of the European Commission.

4 The W7-X Team Alonso [1], Andreeva [2], Baldzuhn [2], Beurskens [2], Beidler [2], Biedermann [2], Blackwell [17], Blanco [1], Bosch [2], Bozhenkov [2], Brakel [2], Burhenn [2], Buttenschön [2], Cappa [1], Czarnetzka [3], Dinklage [2], Endler [2], Estrada [1], Fornal [3], Fuchert [2], Geiger [2], Grulke [2], Hartmann [2], Harris [4], Hirsch [2], Hoefel [2], Jakubowski [2], Klinger [2], Klose [2], Knauer [2], Kocsis [5], König [2], Kornejew [2], Krämer-Flecken [6], Krawczyk [3], Krychowiak [2], Kubkowska [3], Kiazek [7], Langenberg [2], Laqua [2], Laqua [2], Lazerson [8], Maaßberg [2], Marsen [2], Marushchenko [2], Moncada [9,10], Moseev [2], Naujoks [2], Otte [2], Pablant [8], Pasch [2], Pisano [11], Rahbarnia [2], Riße [2], Rummel [2], Schmitz [12], Schröder [2], Stange [2], Stephey [12], Szepesi [5], Trimino-Mora [2], Thomsen [2], Traverso [13], Tsuchiya [14], Turkin [2], Velasco [1], Wauters [15], Werner [2], Wolf [2], Wurden [16], Zhang [2], et al. [1] CIEMAT, Madrid (Spain) [2] Max-Planck-Institut für Plasmaphysik, Garching, Greifswald (Germany) [3] IPPLM, Warsaw (Poland) [4] Oak-Ridge National Laboratory, Oak Ridge, TN (USA) [5] Wigner RCP, Budapest, (Hungary) [6] Forschungszentrum Jülich, Jülich (Germany) [7] Opole Univerisity, Opole (Poland) [8] Princeton Plasma Physics Laboratory, Princeton, NJ (USA) [9] CEA, Cadarache (France) [10] ThermaVIP Ltd., Cadarache (France) [11] University of Cagliari, Cagliari (Italy) [12] University of Wisconsin, Madison, WI (USA) [13] Auburn University, Auburn, AL (USA) [14] National Institute for Fusion Science, Toki (Japan) [15] ERM, Brussels (Belgium) [16] Los Alamos National Laboratory, Los Alamos, NM (USA) [17] Australian National University, Canberra (Australia) Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 4

5 Outline 1. Introduction: stellarators in a nutshell 2. Measuring W7-X Results from the first operation phase of W7-X Diagnostics developments and aspects relevant to steady-state operation 3. Summary contributions to BPXs and open issues for stellarator reactors Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 5

6 I. Introduction What is a stellarator and how does it work? What is the potential of stellarators for a future fusion power plant? Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 6

7 Magnetic Confinement What is a stellarator? How to get to equilibria for toroidal plasmas? Need for rotational transform and flux surfaces Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 7

8 Tokamak vs Stellarator? Two main approaches to twist toroidal magnetic fields: gas engine and diesel of fusion. Tamm Sacharow Spitzer Tokamak Stellarator What is a stellarator? Stellarators: + steady-state, no large plasma currents, - 3D losses/engineering, one generation behind Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 8

9 Tokamak Confinement & symmetry What is a stellarator? E. Noether ( ) If a system has a continuous symmetry property, then there are corresponding quantities whose values are conserved in time. large toroidal currents generate rotational transform Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 9

10 poloidal angle Stellarators What is a stellarator? 3D confinement: helical fields modulate B toroidally 2p BxB and p curvature drifts: q locally trapped particles are quickly lost! 0 0 p/n toroidal angle 2p/N Beidler et al Nucl. Fusion 51, (2011) (2001) Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 10

11 Why Wendelstein7-X? What is a stellarator? The HELIAS* concept and its scientific perspective Assess stellarator optimization: overcome classical stellarator s draw-backs Bring stellarators to maturity: understand hot plasmas in 3D Kleiber, Borchardt et al. How to get from the idea to a power plant? *Nührenberg, Zille, Phys. Lett. A 114, 129 (1986) Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 11

12 Is there a reasonable perspective for the stellarator line for fusion energy? Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 12

13 [T]here are known knowns; there are things we know we know. We also know there are known unknowns; that is to say we know there are some things we do not know. But there are also unknown unknowns there are things we do not know we don't know. Rumsfeld (2002) Schauer, Bykov, et al. Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 13

14 Known unknowns Physics Optimization criteria (W7-X) Divertor loading (W7-X) Confinement scaling (W7-X) Steady-state operation (W7-X) Tritium breeding (Tokamak line) Engineering Coil design (S-DEMO Engineering studies) Blanket design (S-DEMO Engineering studies) Vessel design (S-DEMO Engineering studies) Support design (S-DEMO Engineering studies) Blanket maintenance (S-DEMO Engineering studies) Material issues (Tokamak line) Economics Size optimization (S-DEMO Engineering studies) Operational availability (S-DEMO Engineering studies) Safety Tritium inventory (Tokamak line) Decay heat (Tokamak line) but we have W7-X (and other stellarators), ideas & concepts Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 14

15 How far away is a reactor? Step ladder to fusion power plant How to build a stellarator reactor? Tokamak Stellarator? ISS, Q=10 K. Lackner, Fusion Sci. Technol. 54, 989 (2008) Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 15

16 F Schauer et al., Contrib. Plasma Phys. 50 (2010) 750 Coil design ITER and HSR5 coils (same scale) ITER (TF only) HELIAS 5B Magn.field at plasma axis 5.3 T 5.6 T Maximum magn. field 11.8 T 12.3 T Superconductor Nb 3 Sn Nb 3 Sn ITER toroidal field (TF) coil HSR50a coil #5 Circumference 34.5 m 34.7 m Minimum bend radius 2.0 m 1.63 m Magn. energy per coil 2.3 GJ 3.2 GJ Coil design for reactor scale devices w/ ITER technology Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 16

17 Vessel & support design for HSR Vessel Design studies (Schauer et al, 2012) Support Stresses No central support ring as in W7-X The inter-coil structure consists of bolted panels One or two panels with one or two plates depending on load distribution Panel size ~ m 2 Stresses within allowable limits for stainless steel (1.4429) at 4K Vessel: Double hull structure similar to ITER VV Wall thickness: >60 mm each In-between water & steel shield: 220 mm Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 17

18 Blankets & Maintenance Concepts Idea for maintenance approach Blanket and shield Stresses Blanket & Shield Size of HELIAS 5-B is determined by blanket space requirements Space between coil and plasma: 1.3 m, blanket thickness: 80 cm 400 blanket segments Geometry not yet optimized Using half module symmetry aiming at 40 different segments Flux surfaces are preserved Maintenance One large vertical port per module (5 modules) 4.3x(2.5, 1.8) m² Separation of plasma and outer vessel Remote handling device for separation (KIT) Insertion of maintenance boxes Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 18

19 Where do stellarators stand? Stellarators 3D magnetic confinement facility B generation by coils outside the plasma + steady-state, stable configuration w/o current, reversed shear + no disruptions; radiation collapse slower that tokamak disruptions + high-density operation possible lower p a at given Q - 3D engineering integration and maintenance - concept development one generation behind the tokamak many unknowns About stellarator DEMO control + less effort needed for real time control of current and plasma position + milder instabilities - plasma scenario to be explored (confinement, impurity control) - burning plasma effects unknown - divertor operation and detachment control to be explored Adapted from W. Biel et al., SOFT San Sebastian Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 19

20 II. Measuring W7-X W7-X is the key device for the stellarator line. How does it specifically contribute to research and operation of larger (burning) stellarators and what are the potentials to contribute to developments for burning plasma devices in general? Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 20

21 W7-X: HELIAS en route to a FPP Wendelstein 7-X Basic research: ionized matter under extreme conditions Energy research: confinement of hot plasmas for fusion Mission: bring stellarators to maturity

22 Wendelstein 7-X Greifswald (Germany) operating since HELIAS-type stellarator N f =5, R/a = 5.5m/0.53m 30 m 3 plasma volume heating ~8+7MW (ECRH, NBI) ICRH (~1.2 MW, later upgrades) no DT operation: plasma physics experiment 70 superconducting coils (2.5T) 5 x 2 x 5 non-planar coils 5 x 2 x 2 planar coils Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 22

23 Wendelstein 7-X Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 23

24 Optimization and phys. requrmts. What is a stellarator? Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 24

25 Project time-line Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 25

26 Summary? First experiments on Wendelstein 7-X Goal of the first operation phase [1] : 1. demonstrate the existence of flux-surfaces 2. integral commissioning of a complex machine: magnets, cryostat, heating, diagnostics, control & data acquisition and if there was time to do some experiments 2015 OP 1.1 (13 Wks) Limiter configuration Pulse limit P dt 2 MJ pulse ~ 1 s P < 5 MW T e, T i 3 kev, 1 kev n 0.2 x m -3 < 1.6 % Initial plan: demonstrate operation and do first measurements T.S. Pedersen et al., Nucl Fusion 2015 Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 26

27 Milestone #1: proof of flux surfaces M. Otte Can W7-X (a HELIAS) be built? B 11 is within the allowed margins so far. Andreas DINKLAGE for the W7-X Team Wendelstein 7-X Universität Siegen 16. Jun Page 27

28 W7-X Camera Systems Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 28

29 Milestone #2: First plasma Dec. 10th, 2015 First helium plasma in W7-X was created according plan Full field: B = 2.52 T P ECRH = 1.3 MW pulse = 50 ms First experiments on Wendelstein 7-X First measurements of plasma parameters conducted T e ~ 100eV n e ~ O(few m -3 ) First shot program (sequence of conditioning pulses) w/ steady-state control system conducted on Dec. 11th Can a HELIAS be operated? Yes. Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 29

30 Safe Operation: Camera Systems Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 30

31 W7-X Camera Systems Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 31

32 Milestone #3:First Hydrogen Plasma in W7-X , 15:21: (local time) First experiments on Wendelstein 7-X 60 s later First hydrogen plasma in Wendelstein 7-X total heating power: ~ 2 MW pulse lengths: ~ 250 ms T e ~ 7 kev, T i ~1.2 kev <n> ~ 2 x10 19 m -3 a < 49 cm (V ~ 26 m 3 ) (plasma touched the limiter) P (MW) absorp. T e (kev) T ion (kev) n e dl (10 19 m -2 ) H a (neutrals, a.u.) limiter current (a.u.) 0 time (s) 0.3 W7-X Team Greifswald, Germany 03.Feb.2016

33 n e dl (10 19 m -2 ) (kev) (kev) (MW) Prolongation of discharge duration limiters not overheated even in 2 MJ discharges, 4 MJ per discharge was allowed during the last weeks of operation From 1 s to 6 second discharge shown (1 s 1MW, then 5 s 0.6 MW): First experiments on Wendelstein 7-X P ech T e T i heating power electron temperature ion temperature line density time (s) Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 33

34 Overall performance Confinement for first plasmas not degraded w.r.t. tokamaks Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 34

35 Diagnostics for plasma core studies * to come Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 35

36 Magnetics Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 36

37 Magnetics for long pulses Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 37

38 3D aspects Foldable Diamagnetic Loop to Avoid Thermo-voltages at Connection Plugs Folded loop made of continuous ribbon cable fits through large ports for installation In-vessel assembly test without ECRH stray radiation shield

39 ECRH protection ECRH Protection for Rogowski Coils and Connection Boxes Thin perforated SS tubes for ECRH protection and outgassing and Cu bars for cooling SS tubes via heat conduction Andreas DINKLAGE Andreas Int. DINKLAGE School of Visit Fusion of the Reactor PMU Technology at W7-X Greifswald ERICE, Italy 17. Feb. 03. May Page Page 39 39

40 Steady-state electronic drift compensation Long pulse integration, all currents (PS, bootstrap) are small (W7-X Optimisation) Excellent performance of integrator, well suited for 30 min discharges and low signals in W7-X, max. drift 72 A over 30 min drift & common mode signals eliminated by chopping and numerical processing Drift measured with 100 A wire through Rogowski coil installed on W7-X plasma vessel: drift: 4 A/100s Lowest expected plasma current 5 ka in W7-X A. Werner Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 40

41 n e dl Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 41

42 n e dl for steady-state operation Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 42

43 n e dl for steady-state TEXTOR Dispersion interferometer module Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 43

44 n e dl for steady-state operation principle: 5 mm 10 mm simplified ANSYS model: T-distribution 293. o C heatload deformation Mo 2 cm Mo CuCrZr 270. o C deformation across surface cooling Dl = 50 mm expected analytically for slab: Dl = 40 mm fastened expected: Dl = 2.9 cm * *(282-20) o C = 40 mm Dz = mm analytical slab model delivered Dz 0.17 μm 0.03 mm / division heatload to mirror 90 kw/m 2 device fixed at bottom T_interface=270 o C -> T_max=290 o C -> smooth temperature distribution and deformation Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 44

45 Thomson scattering on W7-X Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 45

46 X-ray Imagings Spectrocopy T i, E r Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 46

47 T i Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 47

48 Heat loads First experiments on Wendelstein 7-X (c) G. Wurden, LANL A Heat load shifts upwards as the n=1 perturbation trim coil currents with a maximum in Module 3, are increased (while holding the phasing fixed). M.Jakubowski, S. Lazerson, et al. Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 48

49 Surveillance for safe operation Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 49

50 Surveillance for safe operation Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 50

51 Thermal loads from radiation Heat load distribution across the first wall at P rad = 10 MW 3D Monte Carlo simulations 10 7 test photons on flux surfaces surface primitives (ANSYS mesh) 30 min. on Linux PC Protective measures: Water cooled SS heat shields, water cooled windows & mirrors Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 51

52 S-DEMO control requirements table Quantity Diagnostics Actuators Interactions Control accuracy Main plasma 2 % -? density Main plasma temperature Plasma position and shape Zeff, impurity composition Fusion power Plasma instabilities Divertor detachment and heat flux control Polarimetry Reflectometry Spectroscopy Neutrons ECE Spectroscopy Neutrons Reflectometry ECE Magnetics behind blanket Spectroscopy U loop FW and Div coolant temperature Reflectometry ECE Reflectometry ECE Spectroscopy Divertor current Gas injection Pellet injection Pumping system Aux. heating Gas injection PF coils CS coils Plasma heating Impurity gas inlet Gas injection Pellet injection Impurity inject. Aux. heating ECRH PF coils Gas injection Pumping system Wall and divertor, temperatures (outgassing) Main plasma density Confinement (beta) FW and Div fluxes, erosion Confinement (beta) q profile beta density Zeff Confinement (beta)? Spatial resolution Control time response a/10 in core s for 10% increase,?? 5% a/20 in edge ndl, ~n (10%) 1 5 particle ~ few s s for 10% decrease n us 5% -? a/10 in core Several s for increase,? 10% a/20 in edge A few ms for decrease T 0, T us a/50 a/100 > 0.1 s (PF? coils)? < 0.1 skin, s (confinement) L/R ~ 10s.. min (feed-forward?) Integral or a/5 1 s? 0.5 Pmax/50 Integral Several s for increase, A few ms for decrease t.d.b.? a/40 (t.b.d.) < 1 ms t.b.d. t.b.d. 10 ms? Adapted from W. Biel et al., SOFT San Sebastian

53 Summary I: W7-X Results W7-X first phase: all technical and scientific objectives successfully achieved [1] safe routine operation of ECRH, cryo-plant, coil system and control/daq demonstrated about 20 diagnostics successfuly commissioned and delivered results allowed the W7-X Team to safely increase technical limits (2 4 MJ) allowed one to change magnetic configurations opened the door for an unexpectedly comprehensive physics program First experiments on Wendelstein 7-X : fundamental research: self-organization high electron temperatures, turbulence assessing a potential path to a fusion power plant: assembly accurate, flux surface, first step to steady-state operation and even more physics topics in view of future operation addressed: confinement/transport, heating and drive of plasma current, tools for exhaust [1] Sunn Pedersen et al., Nucl. Fusion 55 (2015) Andreas DINKLAGE for the W7-X Team Wendelstein 7-X Universität Siegen 16. Jun Page 53

54 Summary II: Potentials Optimized Stellarators: a potential path to a FPP W7-X is to bring the HELIAS line to maturity OP1.1: good flux surfaces, neoclassical effects Physics and technological boundary conditions drive diagnostics Qualify key technologies (modular coils, steady-state operartion, ECRH) Show good plasma confinement & proof optimization Demonstrate high-performance steady-state operation (heating, fuelling, exhaust, impurities, fast-ions, MHD) Diagnostics for S-DEMO? Andreas DINKLAGE Int. School of Fusion Reactor Technology ERICE, Italy 03. May Page 54