- Plasma Control - Stellarator-Heliotron Control

H.Yamada National Institute for Fusion Science, NINS The University of Tokyo

Acknowledgements to T.Akiyama, A.Dinklage, T.Goto, M.Koyabashi, J.Miyazawa, T.Sunn Pedersen, R.Sakamoto, C.Suzuki, N.Tamura the LHD Experiment Group and the W7-X Team

5th IAEA DEMO program WS, 7 May 2018, Daejon, Korea 1 Commissions from W.B. for H.Y.  Plasma control issues for a stellarator DEMO reactor and how to address them.  The emphasis should be on the question what is different (easier?) when considering the control of a stellarator reactor rather than a tokamak reactor  Less demanding equilibrium control, less demanding density control, no current drive, and no disruptions may be assumed for the stellarators, but maybe some additional need for avoiding impurity accumulation. What about power exhaust control in a stellarator reactor?  Use examples from LHD and W7-X (experimental and modelling)  Extrapolate to the real conditions for a stellarator DEMO reactor and formulate some predictions and recommendations.

2 Outline 1. Introduction  Quick review of stellarator-heliotron activity  Leading concepts

2. Specific characteristics of stellarator- heliotron plasmas  such as disruptions, high density, MHD, divertor

3. Control of stellarator-heliotron DEMO reactor

4. Summary

3 World-Wide Stellarator-Heliotron Experiment Programs

 China Japan, EU (Germany, Spain), USA and rising China

4 Diversity of stellarator-heliotron concept Now zoology has converged in 4 major concepts

Heliotron QI (LHD) HELIAS (W7-X)

Advantages and disadvantages co-exist Complementary (and competitive) approach must be promoted QHS QAS (HSX)

Note: QAS is a tokamak-stellarator hybrid concept (needs plasma current) 5 Fueled m He(Pa

) 3 300

Stellarator: -3

m 2 200 Intrinsic capability 19 ne Qfuel 1 100

of steady-state operation with (x10 e 3 n favorable potentials such as 0 0 ) 4 Ti0 Te0 3  Free of current drive (even for QAS) 2

T (keV) T 1 0  Free of current disruptions  thermal quench 2 Total 1

 Very high density operation (MW) ICH RF ECH P 0  MHD instability seems to be benign 0 1000 2000 3000 time (s)  Stable detached divertor operation

 Much less demanding feedback control

How relevant is this statement ?

Take a look of facts

6 Issues related to disruptions and plasma current control Tokamak-Stellarator hybrid: Stabilization of MHD instability by external rotational transform was already demonstrated long time ago.

JIPP T-II W VII-A in 1981 in 1979

 No current with BSC QAS relies on bootstrap current Not small, 4MA in ARIES-CS disruption? could be assumed vacuum Heliotron is accompanied by more gentle than significant BSC (several MA in FFHR) in tokamak ARIES-CS Najmabadi FST 2008 QI and QHS is not accompanied by significant BSC. But complete suppression of plasma current is required in order to secure divertor configuration. 7 Heliotron plasmas are so resilient: - Typical temperature hole (hollow plasma) formation -

C. Suzuki et al., Plasma Phys. Control. Fusion 59, 014009 (2017).

total (perp.) NBI#4,5 (tangential) NBI#1,2,3 ECH

2 τ ≃ 0.2 s Recove keV ry

4×1019 m-3

Prad ≃ 2.4 MW accumulation (Pheat = 9.5 MW)

69Tm TESPEL courtesy of C.Suzuki 8 NOT a vacuum inside the temperature hole

C. Suzuki et al., J. Phys. B: At. Mol. Opt. Phys. 45, 135002 (2012). (a) (b) (c)

× 17 (a) total NGd = 1.98 10 Te ne

4.5 s

(b)

5.1 s

(c)

5.9 s

19 -3 ne ≃ 2×10 m inside the hole 64Gd TESPEL (suggested from FIR interferometer) courtesy of C.Suzuki 9 Temperature hole dynamics induced by fuelling pellets

Pheat Prad

pellet Hb

Target plasma Courtesy of A.Dinklage 10 Temperature hole dynamics induced by fuelling pellets

Pheat Prad

pellet Hb

First Pellet Courtesy of A.Dinklage 11 Temperature hole dynamics induced by fuelling pellets

Pheat Prad

pellet Hb

Second Pellet Courtesy of A.Dinklage 12 Temperature hole dynamics induced by fuelling pellets

Pheat Prad

pellet Hb

Temperature Hole

Courtesy of A.Dinklage 13 Temperature hole dynamics induced by fuelling pellets

Pheat Prad

pellet Hb

Recovery Recovery time scale is longer than t Courtesy of A.Dinklage E 14 Helical systems can be operated in much higher density regime than tokamaks

2 W7-AS has also extended high Greenwald density limit nG J  I p /  a exp density (n >>nGW) a2 B 1  2 I 5   M.Greenwald, PPCF 2002 pR 2 a LHD ) 3 /m 20 1.0 (10 exp e

n M. Greenwald, PPCF 2002 Alcator C DIII PBX M.E.Puiatti, NF 2009 0.1 RFX-mod 0.1 1.0 10 20 3 nGW (10 /m ) Extension of favorable density dependence ISS 04 2.28 0.64 0.610.54 0.84 0.41 tE  0.134a R P ne B 2 / 3 higher fusion reactivity, easier plasma solutions of divertor reduced fast-ion instability, fast-ion loss to walls, & neoclassical ripple loss 15 Discovery of super-dense-core operation  Usually density profile is hollow

 Gas-fueled discharge : hollow density profile n  Density profile in the plasma with an GW highly peaked density profile

Much higher density than a usual gas-fueled plasmas with the higher temperature  Confinement improvement pronounced in the core leads to higher central pressure

Central density reaches 1.11021m-3 at 2.5 T Central pressure 1.5  atmospheric pressure

Note: peaked density profile is due to central fueling

(pellets can penetrate into the core because of low Te) 16 Density limit is determined by power balance in the edge

Sudo 2 0.5 ne  0.25( PBaR /( )) only limits edge density 0 2 30 3.65 m: HD: Gas puff Pellet 3.65 m: HD: Pellet (w/o IDB) Gas puff ) 3.65 m: LID: Pellet (IDB) Sudo 3.65 m: LID: Pellet (IDB) -3 n c 3.75 m: HD: Pellet (IDB) m 20

20 3.85 m: HD: Pellet (IDB) 100eV

1 e (10 / n / Sudo e0

100eV 10 n = 4 n n e e0 c n

0 0 2 4 6 8 101214 0 0 0.5 1 1.5 P (MW) tot n 100eV / n Sudo e c

Density at Te=100 eV is Plasmas with SDC has ne0 reaching Sudo a good index 4ne while large margin against the density limit is secured 17 High-beta state is maintained for 100 tE even in moderately unstable regime b (plasma pressure/magnetic pressure) reaches 5 % “Minimum B” Successful paradigm of MHD since 1950’ LHD magnetic hill

)  unstable against interchange mode % ( 6 b

5 Tokamak 4 t /t Unstable sus E >(%) 3 b Magnetic well LHD experiment has demonstrated < 2 “interchange instability in magnetic hill Magnetic hill is benign” 1 n/m= 1/2 2/3 3/4 1/1 Soft limit is observed, due to saturation in 0 confinement not disruption 0 0.2 0.4 0.6 0.8 1.0  Also W7-AS (magnetic well) demonstrated stable 3.4% plasma 18 Core density collapse limits the central pressure

Time scale of Core Density Collapse (CDC) is several hundreds m-seconds Ballooning mode triggers this collapse 19 Divertor in Stellarator-Heliotron Build-in divertor in LHD Island divertor in W7-X Connection length (m) 100 101 102 103 104 1.0 Divertor legs

0.5

0 Without RMP

(m) Z With RMP (m/n=1/1)

-0.5 Stochastic region

-1.0 2.5 3.0 3.5 4.0 4.5 5.0 R (m)

20 With RMP  Stable sustainment of radiative divertor Without RMP  Radiation collapse due to thermal instability Radiative divertor

19 -3  Stable operation 10 ne (10 m ) around density With RMP limit 3 0 Without RMP 20 1Radiation2 3 4 5  Radiation increase (a.u.) by a factor of ~ 3 10 radiation collapse 2 stablilized ) 0 2 1 2 3 4 5 4  Reduction of divertor power load with 2 by a factor of ~ 10 1 RMP 0 load load (MW/m Divertorpower 1 2 3 4 5 without RMP 0.6 a99 (m)  Plasma shrinks at

RD phase due to (MW) power Radiation 0 radiative energy loss 0 2 4 6 810 0.5 and RMP penetration Line averaged density (1019m-3) 4001Wp (kJ)2 3 4 5  No significant 200 degradation of 0 main plasma 1 2 3 4 5 confinement time (s)

Courtesy of M.Kobayashi 21 Modification of 3D edge radiation structure by RMP : 3D numerical simulation Carbon radiation distribution by EMC3-EIRENE Without RMP Outboard 300 Inboard Outboard side side Without RMP 200 Inboard

MW/m3 2.0x100 100

Poloidal Poloidal angle (deg.) Radiation peak at inboard side 2.0x10-1 0 Outboard 2.0x10-2 Outboard 2.0x10-3 With RMP 300

Z With RMP 200 Inboard

R 100 Poloidal Poloidal angle (deg.)

0 Outboard 0 100 200 300 Toroidal angle (deg.)  Without RMP  Radiation peak appears at inboard side  With RMP  X-point of m/n=1/1 island is selectively cooled Courtesy of M.Kobayashi 22 Full stable detachment achieved

Prad

PECH

• Full stable power detachment on all divertors has been achieved  An order of magnitude less power deposited on divertor surfaces for t > 2.1 s, although the heating power remains constant for 0

 Divertor target Te also drops by an order of magnitude  No drop in energy confinement • Bolometer cameras show >95% radiated power

Courtesy of T.Sunn Pedersen 23 Outward convection of impurities, which is not diagonal transport, generates Impurity Hole Extremely hollow impurity profiles observed in high-ion temperature referred to Impurity Hole 5 10-1 #97011(He),#97030(C,Ne) ITB fomation

e 4 /n I Helium (keV) T (0) 10-2 e 3 e 2 Carbon R =3.6m 2 ax r/a=0.5 -3 (keV) T (keV) 1 10 i 1

Neon T Concentration n Concentration T (0) i (m/s) C outward

0 V 0 0.2 0.4 0.6 0 r (m) inward eff 2.0 )

-3 inward -1 Outward convection 1.5 0 24 68 10 m in spite of negative Er -grad T (keV/m) 17 1.0 i outward Impurity density starts (10 0.5 C

to decrease dramatically n 0 by factor of 10 during ITB 1.6 1.8 2.0 2.2 2.4 2.6 formation time(sec) 24 Understanding of impurity transport is progressing

Courtesy of N.Tamura

Impurities from the outside are Theory for impurity transport: contrary to screened by a friction force in the conventional wisdom, impurity accumulation edge surface layer in high density is not inevitable: regime • Er drops out at high impurity collisionality • Temperature screening possible in stellarators S.L.Newton, J.Plasma Phys. 2017 25 Some remarks on stellartor-heliotron DEMO control Helias reactor ( W7-X) FFHR-d1 (LHD)

 Stellarator-heliotron has advantage of much less control actuator and no need ARIES-CS of current drive, consequently larger engineering Q.  Power exhaust is a common critical issue.  construction and maintenance of complex 3-D structure is serious headache.  radial build to accommodate blanket and shield is a critical issue since the coils are located to plasma closely. 26 Comparison of operational regime of tokamaks and stellarator-helitoron DEMOs tokamak FFHR (Jpn DEMO)

Tokamak Helical Small difference in scaling laws (JA Model 2014) (FFHR-d1A) t IPB98( y ,2) n 0.41 P 0.69 IPB98(y,2) Scaling law ISS04v3 E e t ISS 04 n 0.54 P 0.61 8.5 / 2.43 R / a 14.6 / 2.22 E e 1.65 k — makes significant difference in 5.94 B 4.55 POPCON 12.3 I —  tokamak: high T and low n p  S-H: low T and high n 1.31 H factor 1.0 Courtesy of T.Goto 27 Comparison of operational regime of tokamaks and stellarator-helitoron DEMOs tokamak FFHR (Jpn DEMO)

Tokamak Helical (JA Model 2014) (FFHR-d1A) Contours around operational IPB98(y,2) Scaling law ISS04v3 point  tokamak: very steep 8.5 / 2.43 R / a 14.6 / 2.22  S-H: very gentle 1.65 k — 5.94 B 4.55  in S-H, tiny increase in density 12.3 I — changes fusion output p drastically 1.31 H factor 1.0 Courtesy of T.Goto 28 Effect of fueling depth of pellet injection /13 No inward pinch due to turbulence  Density profile is fixed to be flat no need of profile measurement

❖ Self-burning is possible under wide range of fueling conditions • λ/a> 0.3: self-burning • Fueling depth affect burning properties through density profile changes ❖ Deep fueling contribute to • Minimization of minimum fusion output • Increase of burning efficiency • Minimization of fuel particles • Whereas extremely high pellet velocity is required

Courtesy of R.Sakamoto 29 Operation scenario of heliotron DEMO reactor

Plasma startup by feedback control of the along the preprogramed one. 18 -3 ne=110 m  Pf = 0.3GW Heating power is preprogramed. To avoid radiation collapse,

edge The edge density ne have to be smaller than the density limit “sudo sudo limit” ne , which is determined by the absorbed power Pabs

(Pabs = Pa + Paux - PBr)

Depends on ne, Te profiles, α particle Parameter to be measured：in red profile (neutron profile) Courtesy of T.Akiyama & T.Goto 30 Necessary diagnostic set in S-H DEMO

Parameter Diagnostics Resolution Time resolution Interferometer < 1×1017 m-3 < 10 ms Polarimeter 18 -3 Edge ne profile Reflectmeter < 1×10 m < 100 ms n profile LIDAR Thomson ~ 1×1018 m-3 e < 100 ms Te profile scattering ~ 50 eV a heating power Neutron camera < 100 ms Pbr Spectroscopy Micro-fission camber Neutron flux ~10% < 100 ms Foil activation Ip for QI, QAS Rogowski coil ~ 10 kA 1 s

Under consideration • Detachment control (spectroscopy, Langmuir probe?) • • Equilibrium (saddle loop) (to recover Shafranov shift) for heliotron and QAS

• nHe (spectroscopy?) (He dilution) • nD/nT (TAE or GAM spectroscopy, Neutron camera) • Zeff (spectroscopy?)

Courtesy of T.Akiyama 31 Summary

1. Stellarator-Heliotron (S-H) certainly has a strong appeal to easier control than tokamaks because it is free from plasma current except for bootstrap current. Indeed we have had many encouraging and supportive evidences in LHD and W7-X 2. However, there remain control issues and we must be humble to prospect DEMO. To be easier than tokamak DOES NOT necessarily mean to be technically achievable. 3. Plasma terminating events do exist while they are much more gentle than disruption in tokamak. It should be noted that assessment of impact (released energy, REs, etc.) on DEMO operation must be done carefully. For example, we know disruption is not allowed but we DO NOT know radiation collapse is allowed or not. 4. Control of Shafranov shift in heliotron, and plasma currents in QI and QAS should be cared. 5. Understanding of underlying physics of impurity accumulation and power exhaust is immature yet. Complementary and interactive study with tokamak could accelerate both lines, in particular through 3-D physics.

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