First plasma operation of Wendelstein 7-X
R. C. Wolf on behalf of the W7-X Team*)robert.wolf@ipp.mpg.de*) see author list Bosch et al. Nucl. Fusion 53 (2013) 126001

The optimized stellarator Wendelstein 7-X
Main advantage
 Intrinsically steady-state with benign operational boundaries (w/o disruptions)1.

Closed magnetic surfaces /sufficiently small error fields

2.

Reduced neoclassical transport (thermal plasma); minimization of eff

3.

Confinement of fast ion (in W7-X ~100 keV)4.

MHD stability at finite  (5%)

5.

Equilibrium properties at finite  (5%):
Low Shafranov shift, small bootstrap current

6.

Compatibility of magnetic field and exhaustconcept (in W7-X magnetic island divertor at )7.

Feasible modular coils

Special property of W7-X:
Plasma and magnetic field
as far as possible decoupled

Main objective
 Demonstrate integrated high power, high nTE steady-state plasma operation
R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

2

The optimized stellarator Wendelstein 7-X
Magnetic field
3T
Superconducting coils
70
Cold / total mass425 t / 700 t
Magnetic field energy
600 MJ
Plasma volume
30 m3
Plasma duration
30 minutes

Heating power
10 MW
Maximum heat load
10 MW/m2

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

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The optimized stellarator Wendelstein 7-X
Magnetic field
3T
Superconducting coils
70
Cold / total mass
425 t / 700 t
Magnetic field energy
600 MJ
Plasma volume
30 m3
Plasma duration
30 minutes

Heating power
10 MW
Maximum heat load
10 MW/m2

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26th IAEA Fusion Energy Conference, Kyoto

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The optimized stellarator Wendelstein 7-X• Assembly of basic
device completed in2014• Commissioning
2014 – 2015
• First plasma
operation
10 Dec 2015 until
10 March 2016

Bosch, PD
R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

5Confirmation of flux surfaces and low error fields• Flux surfaces / field line traces measured with electron beam
• Island chain at m/n = 5/6 corresponds to  = 5/6 : Movement of  = 5/6 position
consistent elastic deformation of modular coils when magnetic field is applied• From dependence of central island width of  = ½ configuration on deliberately appliederror field an intrinsic error field of b21  510-6 is deduced
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26th IAEA Fusion Energy Conference, Kyoto

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Contents• Plasma break-down, wall-conditioning and achieved
plasma parameters• First attempt of a global power balance• Confinement and plasma transport• Heating and current drive experiments• Summary and conclusions• Towards steady state operation

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26th IAEA Fusion Energy Conference, Kyoto

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Contents• Plasma break-down, wall-conditioning and achieved
plasma parameters• First attempt of a global power balance• Confinement and plasma transport• Heating and current drive experiments• Summary and conclusions• Towards steady state operation

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

8Plasma generation
Boundary conditions
Helium gas-inlet• Limiter plasma restricted to  P dt  2 MJ
• Six 140 GHz cw-gyrotrons for central
140 GHz ECRH at 2.5T; total power  5 MWPower of4 gyrotrons• 1 week baking in advance @150° C• Start with He w/o GDCNeutral gas pressure(manometer)

Radiation
collapse

First results• Plasma break-down within 10ms

Shine-throughpower• Contamination of wall limits pulse lengthof first discharges to 20ms (automatic stopby sniffer interlock)

Sniffer probe
signal• Hundreds of short ECRH cleaningdischarges (3 days corresponding
to about 4 sec plasma operation)
30s

Marsen, EX/P5-13
R Wolf, 18 October 2016

30s

30s

 discharge length extended to ~50ms
26th IAEA Fusion Energy Conference, Kyoto

9Wall conditioning• Wall conditions improved considerably with cumulated discharge time and GDC• What remained throughout the campaign was the tendency for a radiation collapse
terminating the plasma as wall conditioning deteriorates during the day• The origin of a local neutral gas pressure increase (in module 4) could not be resolved• Eventually, 600 kW / 6 sec discharges were achieved (increasing  P dt to 4 MJ)

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

10Plasma parameters ne dl [1019 m-2]; l  1m
2016-03-08 09:49:52

4

Thomsonscattering

2Single channel dispersion
interferometer

00.00.5

1.0

1.5t (s)Hydrogen plasma
PECRH = 0.6 MW, 6 sec

Hydrogen plasma
PECRH = 4 MW

• At the end of the first W7-X campaign 30 diagnostics were commissioned andprovided data
Pablant, EX/P5-6• Low densities and electron heating by ECRH resulted in Te >> Ti
Langenberg, EX/P5-3• Results in Core Electron Root Confinement (CERC)
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26th IAEA Fusion Energy Conference, Kyoto

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A “typical” hydrogen plasma …

… with mid-plane
manipulator in action

12R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

Contents• Plasma break-down, wall-conditioning and achieved
plasma parameters• First attempt of a global power balance• Confinement and plasma transport• Heating and current drive experiments• Summary and conclusions• Towards steady state operation

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

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Contributions to / assumptions for power balance

Two Bolometer cameras
 Assuming toroidal symmetry
ECRH heating power Transmission efficiency 94%

Wkin from Te(r) and ne(r) Wkin ~ 1.25 Wdia
 Impurity content neglectedIR cameras for limiter loads
 Initial assumptions on asymmetries

Wurden, EX/P5-7
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26th IAEA Fusion Energy Conference, Kyoto

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Fair agreement between heating and loss power• Power balance shows about 10 – 30% of power which is not accounted for(difference increases with increasing heating power)• Assumptions are only approximations• Additional losses could e.g. come from charge exchange processes with backgroundneutrals
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26th IAEA Fusion Energy Conference, Kyoto

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Contents• Plasma break-down, wall-conditioning and achieved
plasma parameters• First attempt of a global power balance• Confinement and plasma transport

Hirsch, EX/4-5• Heating and current drive experiments• Summary and conclusions• Towards steady state operation

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

16Comparison with scaling of energy confinement time

Confinement times during 1st W7-X
campaign

 Best plasmas lie on ISS04-scaling Only 16 days of hydrogen operation
 Bare CuCrZr walls
 Conditioning of wall was still ongoing;
impurity issues??

At low power reduced confinement
 Radiation limit ?
 Density limit ?

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26th IAEA Fusion Energy Conference, Kyoto

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Core electron root confinement (CERC)• Deriving neoclassical transport coefficient from measurements of T- and n-profiles (usingneoclassical transport codes DKES and SFINCS)• Radial electric field from enforcing the ambipolarity condition: e(Er) = Z i(Er)Krämer-Flecken, EX/P5-4• Clear evidence for core electron root confinement (Er > 0): Region of improved
confinement, however, covers only a small fraction of the plasma volume• Confinement shows only weak dependence on optimization parameter eff which isconsistent with -transport regime in the presence of an electric field
R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

18Comparison with scaling of energy confinement time

Confinement times during 1st W7-X
campaign

 Best plasma lie on ISS04-scaling Only 16 days of hydrogen operation
 Bare CuCrZr walls
 Conditioning of wall was still ongoing;
impurity issues

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

19Comparison with scaling of energy confinement timeOptimized confinement time as
predicted for W7-X in the ion-regime(for 1/: eff3/2 dependence)
W7-X CERC plasmas (more like regime, only weak eff dependence) ¼ of volume shows CERC
 ¾ of volume at “ISS04 conditions”

Confinement times during 1st W7-X
campaign

 Best plasma lie on ISS04-scaling Only 16 days of hydrogen operation
 Bare CuCrZr walls
 Conditioning of wall was still ongoing;
impurity issues

Comparison of W7-X and W7-AS
CERC plasmas at similar densities(E/E,ISS04)W7X > (E / E,ISS04)W7AS

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26th IAEA Fusion Energy Conference, Kyoto

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Contents• Plasma break-down, wall-conditioning and achieved
plasma parameters• First attempt of a global power balance• Confinement and plasma transport• Heating and current drive experiments• Summary and conclusions• Towards steady state operation

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

21Demonstration of O2-ECRH• Standard heating scenario for first campaign: 2nd harmonic X-mode• Densities between 1.21020m-3 and 2.4 1020m-3 require O2-heating scheme• Because of unfavourable temperature scaling of neoclassical confinement, optimum
confinement conditions require high plasma density• At low power single pass absorption up to 70% measured at 5 keV comparing ECA withand w/o plasma agrees wellwith theoretical predictions• Multi-pass absorption
scheme gives an overall
absorption value of 95%

Planegraphite
tile

ECRH
beam

Curvedsteel
panel

Moseev, EX/P5-11
R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

22Demonstration of O2-ECRH
Startup with 2gyrotrons in X-mode

4 gyrotronsin O-mode• Plasma start-up in X2-mode• For Te  5 keV simultaneous X2and O2-heating• Finally, sustainment of plasma
only applying O2-heating

Stray radiation (sniffer probes)=> nearly 95% absorption
Central ECE

Xmode

R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

X+ Pure O-mode heating
Ondl23Current drive experiments≈ -10°2MW
ECRH

Crashes incentralECE

Counter-ECCD

Agreement
withpredictedcurrents≈ +10°
Central collapses
by iota change

Co-ECCD• Remember: Low shear -profile with   1• Central Co-ECCD increases • Crashes in central Te triggered by magnetic resonances generated by -change ?
R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

24Summary and conclusions
The objective of Wendelstein 7-X is to demonstrate that the steady-state
stellarator confinement concept fulfils the requirements for the development toa fusion power plant Steady-state operation of a high-power, high performance plasma
The superconducting stellarator Wendelstein 7-X was successfully commissioned,
first plasma experiments were very successful
 At PECRH 4 MW, Te0  8 keV, Ti0  2.2 keV, ne0  4.2∙1019m-3, nedl  3.6∙1019m-2
achieved simultaneously Discharges lasting up to 6 sec (Pdt  4 MJ)
 Integral commissioning including 30 plasma diagnostics and 5 MW of ECRH
 A comprehensive physics programme has been conducted with many interesting
results (about half of the 900 discharges dedicated to physics studies) This forms a good basis for the continuing completion of the device towards fullsteady-state capability
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26th IAEA Fusion Energy Conference, Kyoto

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Many more results reported at this conference
Device commissioning H.-S. Bosch et al., “Final integration, commissioning and start of the Wendelstein 7-X
stellarator operation”, FIP (post deadline)
Heating and confinement
 M. Hirsch et al., “Confinement in Wendelstein 7-X Limiter Plasmas”, EX/4-5
 J. Geiger et al., “Plasma Effects in Full-Field MHD-Equilibrium Calculations for W7-X”,
TH/P1-1
 N. Pablant et al., “Investigation of initial plasma parameters on the Wendelstein 7-X
stellarator using the x-ray imaging crystal spectrometer”, EX/P5-6
 A. Langenberg, “Minerva Bayesian Analysis of X-ray Imaging Spectrometer Data forTemperature and Density Profile Inference at Wendelstein 7-X”, EX/P5-3
 D. Moseev, “Application of the ECRH radiation for plasma diagnosis in Wendelstein 7‐X”,
EX/P5-11
 S. Marsen et al., “First Results from Protective ECRH Diagnostics for Wendelstein 7-X”,
EX/P5-13
 J. Ongena et al., “Physics and applications of ICRH on W7-X”, EX/P5-12
 Y. Kazakov, “ICRH Scenarios for Fast-Ion Generation in Wendelstein 7-X“, TH/P4-22
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26th IAEA Fusion Energy Conference, Kyoto

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Many more results reported at this conference
Plasma transport
 O. Grulke et al., “Transport studies during the first campaign of Wendelstein 7-X ”,
EX/P5-14
 A. Krämer-Flecken et al., “Investigation of turbulence rotation in limiter plasmas at W7-X
with a new installed Poloidal Correlation Reflectometry”, EX/P5-4
Heat load distribution on limiters and error field experiments
 S. Lazerson et al., “Error field measurement, correction and heat flux balancing on
Wendelstein 7-X”, EX/P5-5
 G. Wurden et al., “Limiter observations during W7-X first plasmas”, EX/P5-7
 S. Bozhenkov et al., “Enhancement of W7-X performance by symmetrization of limiterloads with error field correction coils”, EX/P5-8
Plasma edge characterization and plasma wall interaction
 P. Drews et al., “Measurement of the plasma edge profiles using the combined probe on
W7-X ”, EX/P5-9
 F. Effenberg et al., “Numerical investigation of 3-D plasma edge transport and heat fluxes
including impurity effects in Wendelstein 7-X start-up plasmas with EMC3-Eirene”,
TH/P6-11
R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

27Special properties of the W7-X limiter configuration

Affected by …

Lc short compared to divertor phase
Three distinct regions on limiter… rotational transform… cross-field transport

Bozhenkov, EX/P5-8… accuracy of limiter positions… intrinsic error fields and application of error field correction coils
R Wolf, 18 October 2016

26th IAEA Fusion Energy Conference, Kyoto

Lazerson, EX/P5-528Contents• Plasma break-down, wall-conditioning and achieved
plasma parameters• First attempt of a global power balance• Confinement and plasma transport• Heating and current drive experiments• Summary and conclusions• Towards steady state operation

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26th IAEA Fusion Energy Conference, Kyoto

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Towards steady-state operation
2015 / 2016
5 MW
4 MJ
6s
Uncooled
graphitelimiters

CuCrZr surfaces

Steel panels
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26th IAEA Fusion Energy Conference, Kyoto

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Towards steady-state operation
2015 / 2016
5 MW
4 MJ
6s2017 / 2018
10 MW
80 MJ
10 s

Uncooledgraphite
limiters

Uncooled
graphitedivertor

CuCrZr surfaces

Graphite heatshields andbaffles

Ongena, EX/P5-12

Steel panels

Kazakov, TH/P4-12

Steel panels
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Towards steady-state operation
2015 / 2016
5 MW
4 MJ
6s

2017 / 2018
10 MW
80 MJ
10 s

Uncooled
graphite
limiters

Uncooled
graphitedivertor

CuCrZr surfaces

Graphite heatshields andbaffles

Steel panels

Steel panels

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26th IAEA Fusion Energy Conference, Kyoto

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Towards steady-state operation
2015 / 2016
5 MW
4 MJ
6s

2017 / 2018
10 MW
80 MJ
10 s

2020 …
10 MW…20 MW
18 GJ
30 minutes

Water cooled divertorUncooledtargetUncooled
10 MW/m2graphitegraphite
limiters
divertor

Actively cooled
steady state highheat flux divertor- 10 MW/m2 -CuCrZr surfaces

Graphite heatshields andbaffles

Graphite heatshields andbaffles

Steel panels

Steel panels

Steel panels

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26th IAEA Fusion Energy Conference, Kyoto

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Towards steady-state operation
2015 / 2016
5 MW
4 MJ
6s2017 / 2018
10 MW
80 MJ
10 s

2020 …
10 MW…20 MW
18 GJ
30 minutes

Uncooled
graphite limiter

Uncooled
graphitedivertor

Actively cooled
steady state highheat flux divertor- 10 MW/m2 -Increase ofheatingpower

Graphite heatshields andbaffles

Graphite heatshields andbaffles

Tungstenwall…

CuCrZr surfaces

Steel panels
Steel panels
R Wolf, 18 October 2016

Steel panels

26th IAEA Fusion Energy Conference, Kyoto

www.helmholtz.de/fileadmin/user_upload/publikationen/
Helmholtz_Roadmap_2015_
web_korr_150921.pdf
34

Acknowledgements

This work has been carried out within the framework of the EUROfusion
Consortium and has received funding from the Euratom research andtraining programme 2014-2018 under grant agreement No 633053. The
views and opinions expressed herein do not necessarily reﬂect those of
the European Commission.

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26th IAEA Fusion Energy Conference, Kyoto35