Wendelstein 7-X in the European Roadmap to Fusion Electricity

Enrique Ascasíbar for the W7-X Team

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 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Andreas DINKLAGE | Hungarian Plasma Physics and Fusion Technology Workshop | Tengelic, Hungary | 09. Mar. 2015 | Page 2 Outline

1. Introduction and Motivation:Stellarators in the EU Roadmap

2. Operational Phases of W7-X: OP1.1, OP1.2

3. Aspects of SSO

4. Summary

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 3 Wendelstein 7-X, the Engineers’ View mass: 725 t ECRH heating plasma vessel 254 ports 117 diagnostics ports R = 5.6 m OP1.1: 5.3 MW a = 0.5 m 3 Vplasma= 30 m B ≤ 3 T = 5/6 - 5/4

outer vessel

20 planar SC coils 10 divertor units support structure 50 non-planar SC coils plasma

1st ECPD 14.-17. Apr. 2015 4 Europe and W7-X / W7-X and Europe

Introduction and Motivation: Stellarators in the Roadmap http://www.efda.org/wpcms/wp‐content/uploads/2013/01/JG12.356‐web.pdf?5c1bd2

EU Roadmap: Eight Missions: Mission 8 Bring the HELIAS line to maturity Long-term alternative to tokamaks: mitigate risks and enhance synergies: Wendelstein 7-X is an integral part of the European fusion development strategy. Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 5 Why stellarators[1]?

Introduction and Motivation: Stellarators in the Roadmap Stellarators: the main alternative magnetic confinement concept to the tokamak. external coils generate rotational transform: 3D-confinement without plasma current

+ intrinsically steady-state - 3D engineering + no current disruptions - 3D core impurity transport + no current driven instabilities - 3D plasma/fast ion confinement + no need significant current drive high neoclassical losses + no runaway electrons - 3D MCF: one generation behind + operation above Greenwald-limit feasible - divertor concept to be verified

+ lower -particle pressure (for given Pfus) - operation scenarios to be developed

Stellarator Optimization[2]: mitigate 3D losses to pave the way to a Fusion Power Plant

[1]Helander, Rep. Prog. Phys. 77, 0877001 (2014), [2]Nührenberg, Zille, Phys. Lett. A 114, 129 (1986)

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 6 Why Wendelstein 7-X?

Introduction and Motivation: Stellarators in the Roadmap Stellarator Optimization[1]: mitigate 3D losses HELIcal Axis Advanced Stellarator (Neoclassical Optimization) taming locally trapped particles by proper shaping of mod B

75keV D+ W7‐AS W7‐X

75keV D+

© R.Kleiber • proof the concept of stellarator optimization to be a viable path to FPPs • fundamentally new scientific field: lmfp-3D plasma physics  3D impurity transport  fast particles & Alfvénic instabilities

 3D turbulence  high- operation at low i~e  improved confinement modes  new (island) divertor & SOL physics

[1]Nührenberg, Zille, Phys. Lett. A 114, 129 (1986)

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 7 And what is needed for a HELIAS-FPP?

Introduction and Motivation: Stellarators in the Roadmap Engineering Study HELIAS 5-B

electro-mechanical feasibility of HELIAS fusion-power plants

Schauer et al., Fusion Eng. Design 88, 1619 (2013)

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 8 And what is needed for a HELIAS-FPP?

Gaps in dimensionless , *, * parameters ITER‐like shaped tokamaks Wendelstein-line 4 4 Lackner, Fusion Sci Technol. 54, 989 (2008) HELIAS 5‐B 3 DEMO 3 2 * * ITER 2 2  10* 10* 2 log(P*) 1 JET log(P*) 1 /2 W7‐X 100* AUG  W7‐X 0 100* 0 1st phase

/2 2* * */2 W7‐AS 2* * */2 ‐1 ‐1 ‐1 ‐0.5 0 0.5 1 1.5 ‐1 ‐0.5 0 0.5 1 1.5 log(B*) log(B*) From Wendelstein 7-X there is still a large gap to the HELIAS reactor but W7-X is large enough to assess reactor physics aspects.

Andreas DINKLAGE | Hungarian Plasma Physics and Fusion Technology Workshop | Tengelic, Hungary | 09. Mar. 2015 | Page 9 High-level objectives and implications

Mission 8: bring HELIAS to maturity Operational Phases of W7‐X  exploit W7-X to prepare a decision on the next-step HELIAS device: HELIAS-FPP? or HELIAS-BPX?  steady-state, lmfp, high-, (low *, *) plasmas and demonstration of favourable 3D confinement/operation  high-level priorities : address key issues in 3D confinement  safe operation schemes and effects of optimization  density control, impurities, fast particles, island divertor/edge  implement actions along high-level priorities  theory driven exploitation and predictive capabilities  delivery of key components and their operation  preparation of experimental schemes and FPP physics basis specific actions to comply with WP14-18: prepare SSO

+ develop in OP1 feasible island divertor scenarios in a robust environment to mitigate later technical risks with water-cooled in-vessel components (OP2) - increase density - explore magnetic configurations compliant with reliable divertor operation + assess aspects of stellarator optimization, develop interpretative and predictive capabilities

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 10 Timeline OP1.1, OP1.2, OP2

Device commissioning Test divertor HHF divertor (cryostat, fields, vacuum) installation installation

weeks 2014 2015 2016 2017 2018 2019

divertor commissioning

1st plasma C uncooled divertor water‐cooled HHF div. C limiter; pulsed op. plasmas steady‐state plasmas OP 1.1 pulsed operation OP 2 OP 1.2 11 W7-X: first OPs in figures

Operational Phases of W7‐X NC OP 1.1 uncooled carbon limiter PECRH ~ 2 MW (…5MW) Te < 3.5 keV NC 2015 He, (H) gas puff Ti < 0.9 keV surveillance diag. n < 2 x 1019 m-3 13 wks pulse limit: Emax < 2MJ magnetics ISS04 < 0.6% Pulse < 1 s ~ basic n, T, imp. diagnostics NC < 1.6%

OP 1.2(a) uncooled test-divertor (C) PECRH ~ 8 MW H, (D) P H ~ 7 MW 2016 NBI NC pulse limit: E < 80 MJ +profiles MHD (n,T, E ,…) Te < 3.5 keV 29 wks max r NC  < 10s … min +impurity diagnostics Ti < 3 keV Pulse ~ n < 1.6 x 1020 m-3

OP 1.2(b) test scraper elements + PICRH ~ 1.6 MW ISS04 < 1.2% 2017 P < 10 MW  < 3% tot ~ NC 29 wks + blower gun pellet inj. + diagn. upgrades

NC OP 2 actively cooled divertor (CFC) PECRH ~ 10 MW Te < 4.5 keV H NC >2019 steady-state capable PNBI ~ 7 MW or Ti < 4 keV D 20 -3 D, H PNBI ~ 10 MW n < 2.4 x 10 m technical limit: 30 min/10MW PICRH ~ 4 MW SS04 < 2 % P /A < 10 MW/m2 P < 20 MW  < 5 % target tot ~ NC + quasi cw pellet injection + steady-state upgrades

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 12 Minimum goals of OP1.1

OP1.1 uncooled carbon limiter, He, (H), Emax < 2MJ OP1.1 priorities: first plasma operation and integral commissioning

1. Integral commissioning of all systems needed for successful plasma operation 2. Existence of closed flux surfaces all the way to the limiter (at B=2.5 T)

3. Measurement and adequate reduction of B11 field errors 4. Reliable ECRH plasma startup scenario in He 5. Basic ECRH interlocks and safe operation scenarios 6. Basic impurity content monitoring 18 -3 7. Central Te>1 keV at ne>5*10 m in at least 10 discharges in He Plus some more physics: • Electron root transport studies • Scrape-off layer studies • First experiments with ECRH and ECCD • First comparison He vs H (startup, pumping, confinement) Confirmation of optimization goals of W7-X will be done in later operation phases. T.S. Pedersen, EPS2014 & W7-X PC Meeting, Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 13 Closed flux surfaces view into the W7-X vacuum vessel OP1.1

EDICAM video camera system

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 14 Field errors

-4 OP1.1 no magneticB11 = 10 fieldT errors

high-iota configuration standard configuration assessment of small field errors  development of divertor configurations

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 15 Limiter SOL physics

OP1.1 • connection length is short Cross field diffusion rate visibly compared to divertor phase affects heat load patterns • two distinct regions on limiter D (m2/s) 0.5 1.0 2.0

Langmuir probes

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 16 Scrape‐off layer physics with a limiter

• Langmuir probes provide direct

measurements ofpower decay length, λq • Can be done piggy-back on confinement studies (vary density and temperature, reach quasi-steady conditions).

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 1717 Predictive simulations of 1st plasmas 100 OP1.1 Plasma 50 energy,kJ

Te, Ti, keV 2

0.96650

0.9662 edge iota 0.9658

0.9655 1500

1000 © Yu.Turkin bootstrap and 500 total current, A • 500 kW ECRH; 100 ms after burn-through

0 • bootstrap current shielded by the plasma response • strong electron-root feature in the core – high Te 0 0.2 0.4 0.6 0.8

Time/s

Andreas DINKLAGE | Hungarian Plasma Physics and Fusion Technology Workshop | Tengelic, Hungary | 09. Mar. 2015 | Page 18 OP 1.2: the test-divertor unit (TDU) phase

OP1.2 Uncooled but robust TDU: unique possibility to explore aggressively the configuration space

It gives flexibility to react on new insights and technical changes OP2 Guiding principles: 1. increase density 20 -3 target: beyond X2-ECRH cut-off (~1.5…2 x water10 cooledm ) target divertor operation, towards high-nT/high-@low-* …

OP1.1 2. employ configurational flexibility optimization (small Shafranov shift, good NC confinement, small bootstrap, MHD stability…)

→ arrange physics topics along these lines 1) Prepare high-density, high-power steady-state operation 2) Prove and assess elements of stellarator optimization

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 19 Guiding Principle 1: increase the density

OP1.2  towards safe divertor operation towards high‐nT E , high n & high T → high E, high‐ 80% EMC3/EIRENE simulations 90% plasma scenario development detachment in stellarator divertor plasmas

1.0 X2 → O2 0.8 80% EC heating90% 0.6 transition fraction

0.4 0.2 radiation 0.0 2 3 4 5 6 separatrix density (1019 m‐3) … with stable configuration with high radiation fractions

(control: Lc, x‐point location, nus) Y. Feng Geiger et al., PPCF (2015)

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 20 Guiding Principle 2: exploit configuration space

OP1.2 W7-X coils & protection components: systematic explorations & adjustments Geiger et al., EPS 2014 mod B W7‐X coil set (1 of 5 modules) red: high blue: low planar coils trim coil

Rated for 10 MWm‐2

modular coils scraper element Risk of overload on the target plates: Scraper elements: CFC monoblocks able to handle Lore et al., Lumsdaine et al., IEEE Trans. Plasma Science (2014) peak heat fluxes ≤ 15 MWm‐2:to protect overloaded areas intercepting part of the flux Configurational flexibility: qualify scenarios & address physics topics Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 21 Prepare Operation Schemes for W7-X

OP1.2 Efficacy through targeted scenario preparation and validation of theory

TJ‐II W7‐X • Context  Actions to prepare short experimental campaigns in fields of strategic impact:

(%) optimization and impurity transport

C6+  Long-term action towards a validated n physics basis for a stellarator fusion exp.  =0.9) power plant  (V)

• 2014 achievement 1

D  (V) (  Experimental schemes B 1

  to assess in-flux-surface potential [1] variation 1  to investigate the interplay of Simulations of potential variations 1 experimental value (red, TJ-II) and resulting C6+ neoclassical and turbulent transport density variations (W7-X), respectively. For low- 19 -3  validation of non-local effects density and temperature (6x10 m , 1 keV) W7- [2] X predictied scenarios, variations exist but have (ISHPDB) a minor effect e.g. for diagnostics interpretation • 2015 Outlook & Prospects (C6+).  Continue scheme development (impurity transport, turbulence). [1] Alonso et al., EPS2014  Broaden stellarator reactor basis [2] Satake et al., EPS2014  Fuelling and density control Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic assessments.Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 22 Impurities, improved confinement

OP1.2 Improved Confinement Modes: High-Density H-mode transition in W7-AS

20 P=2MW (ms) E 

, 10

/30 ISS95

imp scaling 

0 1 234 n (1020 m‐3) e McCormick et al., PRL (2002) avoidance of impurity accumulation: employ the interplay of transport, sources (SOL), and improved operation modes

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 23 Plasma Operation and Scenario Development in OP 1.2

Research Strategy and its Implementation Safety Diagnostics & Control freedom in OP1.2 to prepare high-power divertor operation … … prepare safe steady-state operation

Heating at moderate densities: small ECCD (some 10kA) for configuration adjustments at high-densities: transition from X2- to O2- heating … provide means for high-density operation / configuration control

Fuelling and Density Control avoid core depletion by 3D transport (thermodiffusion) … qualify fuelling schemes … control core and separatrix density

Divertor operation schemes

→ mitigation of operation risks with water-cooled PFCs (OP2)

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 24 Diagnostics and Components for W7-X

Aspects of SSO Materializing European Expertise for W7-X: Hardware deliveries for OP1.1 • Context Examples for hardware developments for W7-X  Prepare the European exploitation of ICRH[1] W7-X by hardware contributions, support actions and dedicated software developments[4]  Focus on fields of strategic interest for bringing the HELIAS line to maturity • 2014 achievement  Diagnostics and Heating hardware prepared under delivery and installation Video Surveillance[2]  V-band reflectometry, video, imaging software, PHA [3], XICS channel ready for OP1.1.  X-ray diagnostics, ICRH underway  Enhancement projects launched • 2015 Outlook & Prospects [1] Ongena et al., PoP 2014  Operation in OP1.1 [2] Kocsis et al., SOFT2014  Enhancement Projects being [3] Kubkowska et al., EPS-DC2015 implemented [4] König et al., EPS_DC2015  Feasibility studies being continued Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 25 Steady-state heating: ECH on W7-X

Preparing SSO @ W7‐X EC heating for W7-X: routine operation and tool for physics program

• ECRH Program (in OP1.1)  Reliable and routine plasma start-up from day one (incl. break-down)  Reliable and routine plasma heating and 20 -3 current drive with X2 mode (ne<1.2×10 m ).  High density plasma heating with O2 mode 20 -3 20 -3 (1×10 m

Aspects of SSO Option for cw-EC heating in harsh environments: remote-steering

W7-X front steering (left) and [2] • Concept rem. steer. launcher (to scale)   optimized corrugated Input beam waveguide[1]: launcher to transform steered beam outside Vac. the plasma vessel window •Pros  compact & high power density (100 MW/m2)  much enhanced RAMI avoiding

 complicated mechanics of fast / 2 movable mirrors mitre-bend  neutron screening possible wave-guide L=4a (mitre-bend) a • Cons  Reduced steering range (<20°)  Reduced focusing • W7-X antisymmetric  R&D for 1MW cw for OP2 remote-steering image of the launcher  input field  High-field side launch on W7-X for OP 2 [1] Prater (1997), Kasparek (2003), Laqua, Plaum Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 27 OP1.1 is just around the corner...

EUROfusion Call for Participation (Apr. 7th). Missions being decided Three experimental areas: • Scenario development: safe operation and pulse extension • ECH Physics • Scrape-Off Layer Physics

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 28 Overview of diagnostic status

IPC meeting March 2015 29 Summary

Wendelstein 7-X in the EU Roadmap

• achieve steady-state, high-nTitE plasmas • gain predictive capabilites in view of burning 3D plasmas Maturity: demonstrate favorable operation + validated capabilities First Operation Phases (pulsed, lower power): • qualify safe divertor & develop steady-state scenarios • start to address the physics of optimization prepare steady-state operation EUROfusion Research Strategy • address high-priority fields in view of Mission 8 optimization, steady-state/divertor, impurities, fast particles, density control, turbulence 2015: preparatory actions & 1st W7-X plasma operation to qualify steady-state capabilities

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 30 Back-up slides

Andreas DINKLAGE| Conference | Venue | Date | Page 31 Why SOL and λq studies are important OP1.1

•The width of the heat deposition region, λq, scales with 1/Bp in tokamaks , and NOT with machine size, leading to a prediction

of λq ≤ 1 mm for ITER and DEMO (problematic).

MAST • Heuristic model: NSTX Goldston, Nucl. Fusion 52 013009 (2012)  λq~Lc*vD C-Mod  In tokamaks, L is proportional

AUG to 1/Bp leading to Bp scaling DIII‐D  In a stellarator, Lc is not related to B but to the inclination of JET p the divertor respect to the field [mm] (exp.)

q lines   Limiter operation gives data

points at Lc ~30-80 m  Divertor operation will give

data at Lc ~100-500 m

T. Eich et al. PRL 107 2011 Bpol,MP [T]

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 32 Predictive Modelling for W7-X

OP1.2 Navigating the configuration space of optimized stellarators

• Context 5MW O2  goal: steady-state operation of 20 -3 1.5x10 m optimized stellarators @ high-, low-*, low-* mod B on last closed flux surface  qualify safe SS divertor operation with + vacuum edge robust PFCs (uncooled test-divertor) 3 magnetic field • 2014 achievement 2.5 individually  target configuration (core plasma) charged coils 2 low-bootstrap current/good confinement for high-density operation identified coil set for one  OP1.1[2] and OP1.2[3] experiment W7-X module strategies

Specific high-mirror configurations to minimize • 2015 Outlook & Prospects residual bootstrap currents with good confinement  Integration core/SOL plasma properties and edge magnetic field (divertor)[1]  Broaden discharge scenarios port-folio [1] Geiger et al., EPS2014, PPCF 2015  Employ modelling for interpretation [2] Sunn-Pedersen et al., EPS 2014 (OP1.1.) [3] Dinklage et al., IAEA 2014 Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 33 W7-X in the Roadmap Introduction and Motivation: Stellarators in the Roadmap Mission 8: bring the HELIAS line to maturity

• proof optimization: NC, , bs, stability, … • qualify steady‐state, high‐performance operation • gain predictive capabilities

Max-Planck- … Institut für Plasmaphysik EUROfusionStrategic WPS1 benefit for Europe • Flagship science project • Synergies and risk mitigation in an integrated fusion road-map

• Physics and technologicalAndreas leadership DINKLAGE | Visit of thein PMU 3D at W7-X MCF | Greifswald | 17. Feb. 2015 | Page 34 W7-X within EUROfusion: Mission, high-level goals and research objectives Research Strategy and its Implementation Mission 8 in the European Roadmap: Bring the HELIAS line to maturity by the exploration of reactor capabilities of optimized stellarators with W7-X

prepare the decision point: is an intermediate-step device (burning-plasma HELIAS) required? WPS1: identify and implement experimental and theoretical actions in view of M8 Concept: theory driven experimental exploitation • 3D configuration effects and optimization  demonstrate optimization targets: improved 3D confinement, MHD stab. (=5%), beyond NC: 3D turbulence, … • develop plasma scenarios with theory models (and validate them in experiments to arrive at reliable predictive capabilities)

• steady-state operation (high nT, , low , )  qualify heating, power exhaust & safe divertor operation

(towards FPP IVCs, sufficient frad) • steady-state operation elements: safety, control, cw heating (ECRH), diagnostics • demonstrate high-power, high-density steady-state discharges

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 35 W7-X within EUROfusion: high-level goals and research objectives Research Strategy and its Implementation • density control and fuelling  develop concepts for high-density steady-state operation develop fuelling techniques to avoid core density depletion • pellet, gas-puff (, SSMB, NI) fuelling: deposition mechnanisms • preparation of fuelling scenarios, e.g. high-field, low-field side 3D pellet fuelling • 3D impurity transport  develop discharge scenarios without impurity accumulation • develop and validate models for high-Z 3D-trasnport • implement impurity transport related diagnostics (passive spectr., active LBO) • 3D fast particle physics  develop predicitive tools for fast particle confinement in 3D fields • generation (NBI, ICRH), diagnostics and validated models • predictive capabilities for fast particle confinement and collective effects • turbulence, improved confinement modes and isotope effect  develop and explore confinement beyond neoclassical transport

• study the interplay of turbulence , magnetic geometry and Er (TJ-II, LHD, …) • develop predicitve theory for 3D turbulence + synergies w/ M1-7: support actions (ITER, DEMO), preparation of FPP physics basis

Priorities: high-level objectives in view of M8, EU expertise, anticipated impact on W7-X program Europe and W7-X/W7-X and Europe

Introduction and Motivation: Stellarators in the Roadmap

Eight Missions: Mission 8 Science,Bring the HELIASMCF alternatives line to maturity to mitigate risks and synergies: Wendelstein 7-X is an integral part of the European http://www.efda.org/wpcms/wp‐content/uploads/2013/01/JG12.356fusion‐ developmentweb.pdf?5c1bd2 strategy.

Operational Phases of W7‐X

PFC-technology shapes the way to reliable, steady-state, high –nTE operation

C uncooled actively‐cooled C limiter metal divertor ? divertor HHF divertor (CFC) OP1.1 OP1.2 OP2 OP3 … pulsed pulsed operation ?steady‐state test‐ 30 wks 12 wks water‐cooled divertor 30 wks assembly divertor assy. (09/03/15)

974 … 2014 2015 2016 2017 2018 2019 2020 Fig.

Roadmap Mission 8 [email protected] Stellarator Optimisation Burning Plasma Stellarator? adapted from JG12.356‐10c*

Stellarator FPP review point

1996‐2000 2010 2020 2030 2040 2050 ?

Primary target of OP1: preparation for the actively cooled (SSO) divertor Long term goal: Basis for a HELIAS FPP

* from: Fusion Electricity: A roadmap to the realisation of fusion energy (F. Romanelli et al., EFDA, 2012)

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 39 2D/3D MCF

Andreas DINKLAGE | Hungarian Plasma Physics and Fusion Technology Workshop | Tengelic, Hungary | 09. Mar. 2015 | Page 40 Plasma start up: why helium?

OP1.1 empirical finding in WEGA (in line with Heliotron-J, LHD) © M. Preynas 14 He - Power=7 kW 3E+18 H2 - Power=7.6 kW 12 2.5E+18 10 2E+18 8 > Factor of 5 < Factor of 1/2 1.5E+18 6 1E+18 4 Delay time [ms] 5E+17

2 Electron density [m-3]

0 0 0.0E+00 5.0E-05 1.0E-04 1.5E-04 2.0E-04 2.5E-04 0.0E+00 5.0E-05 1.0E-04 1.5E-04 2.0E-04 2.5E-04 Pressure [mbar] Pressure [mbar]

• Breakdown easier in He than H • NBI start-up model: no dissociative processes +lower vibrational cooling[1] • Discharges serve to condition the machine

[1]Gradic et al., Nucl. Fusion 55, 033002 (2015)

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 41 Steady-state heating: ECH options for FPPs

Aspects of SSO

Active Top view beam

2 of the equatorial N‐Port with ECRH‐ports ‐15°/0°/15°

Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 42