Physics of Stellarator Divertors Thomas Sunn Pedersen, for the W7-X Team Director of Stellarator Edge and Divertor Physics with Special Contributions from H

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Max-Planck- 531st Heraeus Seminar, April 2013 Institut für Plasmaphysik Physics of stellarator divertors Thomas Sunn Pedersen, for the W7-X team Director of Stellarator Edge and Divertor Physics with special contributions from H. Hölbe, J. Boscary, T. Eich, Y. Feng, M. Hirsch, R. König

T. Sunn Pedersen 531st Heraeus Seminar 1

Outline

• Intro: 2D (tokamak) divertors and 3D (stellarator) divertors • LHD helical divertor versus Wendelstein-line island divertor • Scrape-off layer widths in tokamaks and stellarators • Experience from W7-AS divertor operation • Status report on the W7-X divertor(s) • Scraper element prototype • Summary

531st Heraeus Seminar 2 Limiter vs 2D divertor vs 3D divertor: schematic

Limiter plasma Single null Double null Island divertor divertor divertor

531st Heraeus Seminar 3 The tokamak divertor vs a stellarator island divertor

2D vs 3D - experiments W7-AS

531st Heraeus Seminar 4 Stellarator island divertor principle

From J. Kißlinger, W7-X divertor design review

531st Heraeus Seminar 5 The W7-X edge topology

• “Standard configuration”: Edge iota=1=5/5=m/n • Island chain consists of five independent island bundles • “High iota”: edge iota=5/4. • Island chain is one long bundle • “Low iota”: edge iota=5/6 • Island chain is one long bundle • In all three cases the magnetic shear is low and the island are large and can be diverted From J. Kisslinger, W7-X divertor design review

531st Heraeus Seminar 6 The W7-X divertor 3D layout

Ten identical, discrete divertor units, each aligned to the local magnetic field (4-7 degree inclination angle in W7-X) Simple field line diffusion model (Kisslinger) gives maximal loads of 2 10 MW/m From J. Kisslinger, W7-X divertor design review

531st Heraeus Seminar 7 LHD edge topology and helical divertor

divertor plate ergodic layer

LCFS

divertor leg

• Large shear makes it difficult to make large islands • But it allows overlap of islands with different low-order resonances è stochastic layer is formed easily

• Due to the existence of this stochastic layer, the quantity L|| varies strongly from field line to field line • Field lines are diverged “naturally” by helical coils via an “edge surface layer” to a helically continuous divertor relatively far away from the plasma

N. Ohyabu et al. Nuclear Fusion 34 p. 387 (1994)

531st Heraeus Seminar 8 Background: tokamak edge and divertor physics

• Large amount of data available from machines with a spread in size, shape, and B-field strength • Potentially problematic projections to ITER and a tokamak reactor: • Wetted area (next slides) • ELM heat loads on divertors (not mentioned further in this talk) • Impurity accumulation in ELM-free H-mode • Identification of various attractive regimes that might solve these problems: • Detachment, radiating mantle (solve wetted area problem) • EDA H-mode, Improved L-mode, RMP’s for ELM suppression (solve ELM problem without creating impurity accumulation ) • Uncertainty in how well these attractive regimes are ITER relevant

531st Heraeus Seminar 9 Background: tokamak edge and divertor physics

531st Heraeus Seminar 10 Empirical extrapolations of λq to ITER (T. Eich et al. PRL 107 2011)

−1.19 −0.8±0.1 1.05±0.2 0.1±0.1 0±0.1 λq (mm) = (0.63± 0.08)× Bpol,MP λq (mm) = (0.7± 0.2)⋅ Btor ⋅ q95 ⋅ PSOL ⋅ Rgeo

MAST NSTX C-Mod .) .) .) .) AUG exp

exp DIII-D JET C-Mod [mm] ( [mm] q [mm] ( [mm] q λ AUG λ DIII-D JET

λ [mm] (regr.) Bpol,MP [T] q

• Extrapolations to ITER (Bpol=1.18 T) give rather robustly λq,ITER ≤ 1mm • Not a “carbon PFC” effect – seen in all-metal machines as well Why does Bp determine λq in a tokamak?

• Heuristic model (R. Goldston, Nuclear Fusion 52 013009 (2012)) • Magnetic drifts cause cross field transport

• Plasma flows along the B-field at v=cs/2 towards the divertor • My quick and dirty version of Goldston’s model:

v v 2T / eBR " ∇B+Rc = D = $ 2T 2L m Tm L L v || i 4 i || 4 || T # ⇒ λq = ∇B+Rcτ || = = = ρi τ || = 2L|| / $ eBR T eB R R m i %$ • Note the linear proportionality between L|| and λq

• But where does Bp come in then? Why does Bp determine λq in a tokamak?

" L|| λq = 4ρi $ R $ miT a Bt miT 4a # ⇒ λq = 4 = Bt eB RB eB R L ≈ a $ p p || B $ p % Why does Bp determine λq in a tokamak?

" L|| λq = 4ρi $ R $ miT a Bt miT 4a # ⇒ λq = 4 = Bt eB RB eB R L ≈ a $ p p || B $ p %

• Main scaling is with Bp, as seen in the data. • T at the separatrix does not differ substantially between the tokamaks or at

different heating powers in the study (“Tsep is clamped to 50 eV due to power balance”)

• Even though ITER may have Tsep≈200 eV, that just brings λq up from 1 to 2 mm relative to Tsep≈50 eV Status: stellarator edge and divertor physics

• Divertor data only available from a few high-performance machines, first and foremost: • W7-AS (until 2002) • LHD (currently active area of research) • Some optimism regarding wetted area • Theoretical predictions about impurity accumulation are troubling: • Ambipolarity constraint leads to inward pointing E-field (ion root) • Electric field causes inward transport of high-Z impurities • Need for avoiding edge impurity sources • Identification of various attractive regimes that might solve these problems: • W7-AS detachment • W7-AS High Density H-mode (HDH-mode) • Uncertainty in how HDH will project to W7-X (and to a stellarator reactor)

531st Heraeus Seminar 15 Status: stellarator edge and divertor physics

531st Heraeus Seminar 16 The W7-X divertor 3D layout

531st Heraeus Seminar 17 Long connection length may save us

• An important difference to the tokamak divertor is that a very long connection length is possible • The long connection length comes from the fact that the local rotational transform inside the island is very low • Illustrated here for CNT with visualized field lines using an electron beam, originally near 100 eV, in neutral gas

531st Heraeus Seminar 18 λq is not determined by Bp in a stellarator

• For the picture in CNT (at least 3*7=21 transits) we get L||=21*2πR=120R (R=0.3 m in CNT)

L λ = v τ = 4ρ || > 480ρ q ∇B+Rc || i R i

• And we have seemingly broken the correlation between Bp and λq • In W7-X, we would achieve L λ = v τ = 4ρ || ≥100ρ ~ 4 cm q ∇B+Rc || i R i Status: stellarator edge and divertor physics

• Divertor data only available from a few high-performance machines, first and foremost: • W7-AS (until 2000) • LHD (currently active area of research) • Some optimism regarding wetted area • Theoretical predictions about impurity accumulation are troubling: • Ambipolarity constraint leads to inward pointing E-field (ion root) • Electric field causes inward transport of high-Z impurities • Need for avoiding edge impurity sources • Identification of various attractive regimes that might solve these problems: • W7-AS detachment (also observed in LHD) • W7-AS High Density H-mode (HDH-mode) • Uncertainty in how HDH will project to W7-X (and to a stellarator reactor)

531st Heraeus Seminar 20 Can detachment be achieved in a stellarator? In W7-X? • Detachment is favorable for tokamaks and divertors alike, since it reduces the divertor heat loads by a large factor (~10) • In W7-AS the plasma detached from the divertor, ie. glowing ‘clouds’ of recombining cold plasma appeared above the divertor plates • Occurs at higher densities: No Greenwald density limit in stellarators – access to high density and detachment should be possible in W7-X and a stellarator reactor

531st Heraeus Seminar 21 Favorable operating mode in W7-AS: HDH High-Density H-mode observed near the end of the W7-AS program

• Plasma density high and stable. • Good energy confinement. • Impurity density low and steady due to: • Reduced impurity confinement • Reduced impurity influx • Basic physics of HDH not fully understood; • Will it be reached in W7-X? • Is it a low-temperature phenomenon? • (resistive-ballooning modes?) • Is it similar to the C-Mod EDA mode?

531st Heraeus Seminar 22 Outline

531st Heraeus Seminar 23 First divertor: Test Divertor Unit (TDU)

Baffle modules

TDU Vertical modules

TDU Horizontal modules

• 25m2 TDU: solid graphite tiles, no active cooling • Will be installed during the first operation phase of W7-X (OP1) • Same surface geometry as the steady state water-cooled divertor, also known as the High Toroidal Heat Flux divertor (HHF) divertor closure • No power flux limit per se – ablation may occur once a limit of 50-100 MJ/m2 has been reached • Purpose: development of discharge scenarios that do not exceed 10 MW/m2 heat flux at full power 531st Heraeus Seminar 24 TDU Manufacturing status: Complete

• TDU has been manufactured and delivered; installation has not yet started

531st Heraeus Seminar 25 High Heat Flux (steady state water-cooled) divertor

50 - 61.5 TM1H-TM4H CFC 8 HHF-modules Adjustable Frame Unit

Manifold TM5H-TM6H Clamped LHF-modules 19 Ø 9 CuCrZr Pumping Slid TM7H-TM9H HHF-modules

TM1V-TM4V Target Elements HHF-modules High power Low power Module 10 MW/m2 1 MW/m2 Total area 19 m2 6 m2 Target units (10) Target modules 100 20 Target elements 890 376 250 end of 2013 Plasma facing mat. CFC Graphite

531st Heraeus Seminar 26 Testing of the HHF divertor elements

• Prototypes of the HHF divertor elements have been tested extensively Upside down • Elements withstand 10000 cycles of target showing 10 MW/m2 applied for 10 seconds Front targets CuCrZr cooling structure and • Front target (at pumping gap) can front target “only” take 3 MW/m2 • Front targets may receive stronger loads than that during a transient phase, due to bootstrap current evolution • A so-called “scraper element” has been proposed to prevent these overloads

531st Heraeus Seminar 27 Scraper elements

• Scraper does eliminate overload on the target tips (front tiles) • The overload scenarios are only relevant for OP2 (starting 2019) • But we can simulate the OP2 overload scenarios already in OP1 by seeking out special configurations (see H. Hölbe, poster session)

531st Heraeus Seminar 28 Conclusions

• Stellarator divertors share many similarities with tokamak divertors • World database is much smaller – less confidence in projecting to next step devices • There may be some special advantages to the island divertor concept – in particular heat loads • W7-X will deliver important data on island divertor operation and performance • TDU divertor is manufactured • HHF divertor is in series production, with excellent test results • Please visit the W7-X posters also: (Bozhenkov, Drewelow, Hölbe, Preynas, Rodatos, Warmer)

531st Heraeus Seminar, Bad Honnef, 2013 29 Extra slides

• Why do ST’s fit the simple Goldston-like theory? • More on Goldston’s theory extended to stellarators • Magnetic equilibrium diagnostics on W7-X • Video surveillance diagnostic on W7-X

531st Heraeus Seminar, Bad Honnef, 2013 30 Why do ST’s also fit the trend?

miT 4a λq = eBp R • Many tokamaks have 4a/R~1.3

• But MAST has much larger 4a/R~ 4… shouldn’t it have a much wider λq ?

• Maybe not: The L|| on the outboard side is comparably short since the pitch angle on the outboard side is much steeper than in ‘regular’ tokamaks

Courtesy of NSTX group An ELM from MAST λq is not determined by Bp in a stellarator

• For the picture in CNT (at least 3*7=21 transits) we get L||=21*2πR=120R (R=0.3 m in CNT)

L λ = v τ = 4ρ || > 480ρ q ∇B+Rc || i R i

• And we have seemingly broken the correlation between Bp and λq • In W7-X, we would achieve L λ = v τ = 4ρ || ≥100ρ ~ 4 cm q ∇B+Rc || i R i • But the logic of Goldston’s calculation is broken: When you go around the torus poloidally and toroidally, the magnetic drifts will tend to average out.

• The λq calculation becomes much more complicated in this case. More work is needed. But the situation should be better λq caused by ExB drifts (blobs)?

• What about ExB drift in the scrape-off layer? • Especially in L-mode, blobs are observed in the scrape-off layer and cause - perhaps even dominate - cross-field transport • How can Goldston’s formula be so successful if so?

• Because of quasineutrality, |eφ|/Te is order unity and that provides a connection between ExB and magnetic drifts:

2T v = v = ∇B+Rc D eBR

E Δϕeddy ε1T / e ε1 T vExB = = = = B Breddy Bε2 R ε2 eBR

• Similar scaling, similar amplitude (?) would lead to similar λq – again, more work is needed Magnetic diagnostics on W7-X

• Diamagnetic loops (stored energy) • Rogowski coils (toroidal plasma current) • Segmented Rogowski coils (toroidal plasma current distribution) • Mirnov coils (MHD oscillations)

Fachbeirat meeting, May 2-4 2012 34 Foldable Diamagnetic Loop

Long-pulse operation requires elimination of thermal voltages (would lead to large integration errors) Foldable concept avoids segmentation of loop Test loop shown does not have ECRH protection

Fachbeirat meeting, May 2-4 2012 35 Rogowski Coils

Rogowski coils → measure the total toroidal electric current Segmented Rogowski coils → measure the poloidal distribution of the toroidal electric current

Location → outside and inside the plasma vessel

Fachbeirat meeting, May 2-4 2012 36 ECRH protection for Rogowski coils

Fachbeirat meeting, May 2-4 2012 37 Video Diagnostic In-vessel View

View from AEQ Port Opposite AEQ-port

38 Fachbeirat meeting, May 2-4 2012 Toroidally Viewing Video Diagnostics

Objectives: - plasma monitoring, - detect excessive heat load on plasma facing-components - detect hot spots - determine plasma shape, edge structure - not a spectroscopic system à survey diagnostic

Should be straight-forward…?

Fachbeirat meeting, May 2-4 2012 ECRH, plasma radiation, plasma coatings….

• Water cooled front plate with pinhole shields plasma and ECRH radiation

• H2 gas flow through pin hole prevents coatings of optical components • Extra ECRH absorbed using ITO coating of mirror

Fachbeirat meeting, May 2-4 2012