- Plasma Control - Stellarator-Heliotron Control

Transcription

1 - 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 218, Daejon, Korea 1

2 2 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.

3 Outline 1. Introduction Quick review of stellarator-heliotron activity Leading concepts 2. Specific characteristics of stellaratorheliotron plasmas such as disruptions, high density, MHD, divertor 3. Control of stellarator-heliotron DEMO reactor 4. Summary 3

4 4 World-Wide Stellarator-Heliotron Experiment Programs China Japan, EU (Germany, Spain), USA and rising China

5 Diversity of stellarator-heliotron concept Now zoology has converged in 4 major concepts Heliotron (LHD) QI HELIAS (W7-X) Advantages and disadvantages co-exist Complementary (and competitive) approach must be promoted QHS (HSX) QAS Note: QAS is a tokamak-stellarator hybrid concept (needs plasma current) 5

6 6 Stellarator: 3 Intrinsic capability 2 1 of steady-state operation with favorable potentials such as Free of current drive (even for QAS) Free of current disruptions thermal quench Very high density operation MHD instability seems to be benign Stable detached divertor operation Much less demanding feedback control How relevant is this statement? Take a look of facts n e (x1 19 m -3 ) T (kev) P RF (MW) n e 4 T i T e 1 Q fuel time (s) ECH Total ICH Fueled He(Pa m 3 )

7 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 in 1981 W VII-A in 1979 with BSC vacuum ARIES-CS Najmabadi FST 28 QAS relies on bootstrap current Not small, 4MA in ARIES-CS Heliotron is accompanied by significant BSC (several MA in FFHR) No current disruption? could be assumed more gentle than in tokamak QI and QHS is not accompanied by significant BSC. But complete suppression of plasma current is required in order to secure divertor configuration. 7

8 Heliotron plasmas are so resilient: - Typical temperature hole (hollow plasma) formation - C. Suzuki et al., Plasma Phys. Control. Fusion 59, 149 (217). ECH total (perp.) NBI#4,5 (tangential) NBI#1,2,3 2 kev τ.2 s Recove ry m -3 accumulation P rad 2.4 MW (P heat = 9.5 MW) 69Tm TESPEL courtesy of C.Suzuki 8

9 NOT a vacuum inside the temperature hole C. Suzuki et al., J. Phys. B: At. Mol. Opt. Phys. 45, 1352 (212). (a) (b) (c) total N Gd = T e (a) n e 4.5 s (b) 5.1 s (c) 5.9 s 64Gd TESPEL n e m -3 inside the hole (suggested from FIR interferometer) courtesy of C.Suzuki 9

10 Temperature hole dynamics induced by fuelling pellets P rad P heat H b pellet Courtesy of A.Dinklage Target plasma 1

11 Temperature hole dynamics induced by fuelling pellets P rad P heat H b pellet First Pellet Courtesy of A.Dinklage 11

12 Temperature hole dynamics induced by fuelling pellets P rad P heat H b pellet Courtesy of A.Dinklage Second Pellet 12

13 Temperature hole dynamics induced by fuelling pellets P rad P heat H b pellet Temperature Hole Courtesy of A.Dinklage 13

14 Temperature hole dynamics induced by fuelling pellets P rad P heat H b pellet Recovery Courtesy of A.Dinklage Recovery time scale is longer than t E 14

15 Helical systems can be operated in much higher density regime than tokamaks Greenwald density limit I p a B 5 R 1 a LHD n J I / a G M.Greenwald, PPCF 22 p 2 W7-AS has also extended high density (n exp >>n GW ) n e exp (1 2 /m 3 ) n GW (1 2 /m 3 ) M. Greenwald, PPCF 22 Alcator C DIII PBX M.E.Puiatti, NF 29 RFX-mod Extension of favorable density dependence ISS t E.134a R P ne B / 3 higher fusion reactivity, easier plasma solutions of divertor reduced fast-ion instability, fast-ion loss to walls, & neoclassical ripple loss 15

16 16 Discovery of super-dense-core operation Usually density profile is hollow Gas-fueled discharge : hollow density profile Density profile in the plasma with an highly peaked density profile n GW 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 m -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 T e )

17 17 Density limit is determined by power balance in the edge Sudo e 2.5 n.25( PB /( a R)) only limits edge density n e 1eV (1 2 m -3 ) 2 1 Pellet Gas puff n c Sudo n e / n e 1eV m: HD: Gas puff 3.65 m: HD: Pellet (w/o IDB) 3.65 m: LID: Pellet (IDB) 3.75 m: HD: Pellet (IDB) 3.85 m: HD: Pellet (IDB) Sudo n = 4 e n c P tot (MW) Density at T e =1 ev is a good index n 1eV Sudo / n e c Plasmas with SDC has n e reaching 4n e Sudo while large margin against the density limit is secured

18 High-beta state is maintained for 1 t E even in moderately unstable regime b (plasma pressure/magnetic pressure) reaches 5 % b (%) LHD Minimum B Successful paradigm of MHD since 195 magnetic hill unstable against interchange mode 6 Tokamak t sus /t E LHD experiment has demonstrated interchange instability in magnetic hill is benign Soft limit is observed, due to saturation in confinement not disruption <b> (%) Magnetic well Magnetic hill n/m= 1/2 Also W7-AS (magnetic well) demonstrated stable 3.4% plasma Unstable 2/3 3/4 1/

19 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

20 2 Build-in divertor in LHD Divertor in Stellarator-Heliotron Connection length (m) Divertor legs.5 Island divertor in W7-X Z (m) Without RMP With RMP (m/n=1/1) Stochastic region R (m)

21 Divertor power load (MW/m 2 ) 1 2 n e (1 With RMP Stable sustainment of radiative divertor Without RMP Radiation collapse due to thermal instability 19 m -3 ) Radiative divertor Without RMP With RMP 2 1Radiation (a.u.) W p (kj) time (s) Stable operation around density limit Radiation increase by a factor of ~ 3 Reduction of divertor power load by a factor of ~ 1 Plasma shrinks at RD phase due to radiative energy loss and RMP penetration a 99 (m) No significant degradation of main plasma confinement Radiation power (MW) radiation collapse stablilized with RMP without RMP Line averaged density (1 19 m -3 ) Courtesy of M.Kobayashi 21

22 Modification of 3D edge radiation structure by RMP : 3D numerical simulation Carbon radiation distribution by EMC3-EIRENE Inboard side MW/m 3 2.x1 2.x1-1 2.x1-2 2.x1-3 Z R Without RMP With RMP Outboard side Poloidal angle (deg.) Poloidal angle (deg.) Without RMP Radiation peak at inboard side With RMP Toroidal angle (deg.) Without RMP Radiation peak appears at inboard side With RMP X-point of m/n=1/1 island is selectively cooled Outboard Inboard Outboard Outboard Inboard Outboard Courtesy of M.Kobayashi 22

23 Full stable detachment achieved P rad P ECH 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 <t<4 s Divertor target T e also drops by an order of magnitude No drop in energy confinement Bolometer cameras show >95% radiated power Courtesy of T.Sunn Pedersen 23

24 Concentration n I /n e 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 #9711(He),#973(C,Ne) Helium Carbon Neon r eff (m) Outward convection in spite of negative E r Impurity density starts to decrease dramatically by factor of 1 during ITB formation T i (kev) T e (kev) n C (1 17 m -3 ) T e () inward T i () outward ITB fomation V C (m/s) time(sec) 2 1 R ax =3.6m r/a=.5 outward inward -grad T i (kev/m) 24

25 Understanding of impurity transport is progressing Courtesy of N.Tamura Theory for impurity transport: contrary to conventional wisdom, impurity accumulation is not inevitable: Er drops out at high impurity collisionality Temperature screening possible in stellarators S.L.Newton, J.Plasma Phys. 217 Impurities from the outside are screened by a friction force in the edge surface layer in high density regime 25

26 Some remarks on stellartor-heliotron DEMO control FFHR-d1 ( LHD) Helias reactor ( W7-X) Stellarator-heliotron has advantage of much less control actuator and no need 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. ARIES-CS 26

27 Comparison of operational regime of tokamaks and stellarator-helitoron DEMOs tokamak (Jpn DEMO) FFHR Tokamak (JA Model 214) Helical (FFHR-d1A) IPB98(y,2) Scaling law ISS4v3 8.5 / 2.43 R / a 14.6 / k 5.94 B I p 1.31 H factor 1. Small difference in scaling laws t t n P IPB98( y,2) E e n P ISS E e makes significant difference in POPCON tokamak: high T and low n S-H: low T and high n Courtesy of T.Goto 27

28 Comparison of operational regime of tokamaks and stellarator-helitoron DEMOs tokamak (Jpn DEMO) FFHR Tokamak (JA Model 214) Helical (FFHR-d1A) IPB98(y,2) Scaling law ISS4v3 8.5 / 2.43 R / a 14.6 / k 5.94 B I p 1.31 H factor 1. Contours around operational point tokamak: very steep S-H: very gentle in S-H, tiny increase in density changes fusion output drastically Courtesy of T.Goto 28

29 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>.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

30 Operation scenario of heliotron DEMO reactor Plasma startup by feedback control of the along the preprogramed one. n e = m -3 P f =.3GW Heating power is preprogramed. To avoid radiation collapse, The edge density n e edge have to be smaller than the density limit sudo limit n e sudo, which is determined by the absorbed power P abs (P abs = P a + P aux - P Br ) Parameter to be measured:in red Courtesy of T.Akiyama & T.Goto Depends on n e, T e profiles, α particle profile (neutron profile) 3

31 Necessary diagnostic set in S-H DEMO Parameter Diagnostics Resolution Time resolution Interferometer Polarimeter < m -3 < 1 ms Edge n e profile Reflectmeter < m -3 < 1 ms n e profile LIDAR Thomson ~ m -3 T e profile scattering ~ 5 ev < 1 ms a heating power Neutron camera < 1 ms P br Spectroscopy Neutron flux Micro-fission camber Foil activation ~ 1% < 1 ms I p for QI, QAS Rogowski coil ~ 1 ka 1 s Under consideration Detachment control (spectroscopy, Langmuir probe?) <Plasma current (Rogowski coil) (to avoid minor collapse: not disruption)> Equilibrium (saddle loop) (to recover Shafranov shift) for heliotron and QAS n He (spectroscopy?) (He dilution) n D /n T (TAE or GAM spectroscopy, Neutron camera) Z eff (spectroscopy?) Courtesy of T.Akiyama 31

32 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. 32