A possible breakthrough of power handling by plasma shaping in tokamak

Transcription

1 8 th IAEA-TM SSO, 2015, May Nara A possible breakthrough of power handling by plasma shaping in tokamak M. Kikuchi1, A. Fasoli2, T. Takizuka3, P. Diamond4, S. Medvedev5, Y.Wu6, X. Duan7, Y. Kishimoto8, K. Hanada9, M.J. Pueschel10, D. Told11, M. Furukawa12, L. Villard2, O. Sauter2, S. Coda2, B. Duval2, S. Brunner2, H. Reimerdes2, G. Merlo2, J. Jiang6, M. Wang6, M. Ni6, D. Chen6, H. Du6, W. Duan6, Y Hou6, L. Yan7, X. Song7, G. Zheng7, J. Liu7, A. Ivanov5, A. Martynov5, Y. Poshekhonov5, K. Mishra9, A. Fujisawa9, K. Nakamura9, H. Zushi9, K. Nagasaki8, K. Imadera8, Y. Ueda3, K. Kawashima1, K. Shimizu1, T. Ozeki1, H. Urano1, M. Honda1, T. Ando13, M. Kuriyama13, X. Xu14, P. Zhu15, S. Woodruff16 1Japan Atomic Energy Agency, Japan, 2CRPP-EPFL, Switzerland, 3Osaka University, Japan, 4UCSD, USA, 5Keldysh Institute of Applied Mathematics, Russia, 6Southwestern Institute of Physics, China, 6Institute of Nuclear Energy Safety Technology, CAS,China, 7Southwestern Institute of Physics, China, 8Kyoto University, Japan, 9Kyushu University, Japan, 10University of Wisconsin-Madison, USA, 11UCSD, USA, 12Tottori University, Japan, 12Max Planck Institute fur Plasma Physics, Germany, 13Retired, 14LLNL, US, 15University of Science and Technology of China, China, 16Woodruff S. Inc Our contributions to this IAEA TM SSO, others are posters 1 st NTT WS held just before IAEA TM Demo at Hefei 1

2 IAEA TM SSO (2010) in Vienna R Goldston gave an interesting talk. He is a person who can smell something (Heuristic drift-based model of the power scrape-off width in H- mode tokamaks) alerting ELM is not only a big problem but also inter-elm heat is quite serious. -> Since then, I have been thinking how to answer his question. Do we have solution? M. Kikuchi, WCI symposium (invited) 2012 M. Kikuchi, US-EU TTF plenary2013 M. Kikuchi, APTWG plenary 2013 S. Medvedev, IAEA FEC St Petersburg Post dead line M. Kikuchi, Festiva de Theory 2015 (Invited) M. Kikuchi, IAEA TM SSO (2015) Invited D. Chen, This coference G. Merlo, This conference This has been published in NF 2012 and has highest citations among papers published in

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4 M. Kikuchi, M. Azumi 1. Power handling issue in standard tokamak 7 order difference Joffrin Seki 4

5 Present Fusion power handling scenario is very challenging Surface / Volume ratio is small in Fusion but large in Fission 1000 Heat Flux (MW/m 2 ) RadiaFon heat flux on the Sun Fusion Divertor (even with RRC) Fusion 1 st wall w/o RRC Fission ~1MW/m 2 Fossil ~0.3MW/m RRC=Remote RadiaFve Cooling 10-3 s Duration 1 year High thermal efficiency may be possible only at low heat flux!! 5

6 Any energy system (Fusion) must have reliable heat exhaust scenario Tokamak configuration is optimized for good confinement, but not for power handling. [1] D-shape is good (MHD) for high pedestal pressure with H-mode (ETB), leading to large ΔW loss during ELM. Temporary measure : RMP, Pellet pacing/smbi [2] D-shape leads to X-point toward small R region. This makes power handling more difficult. Temporary measure : Snow flake, Super X 6

7 1.1 DEMO power exhaust scenario (>20 years before) For 3GW, Q=50 system 660 MW heat out of plasma center q=70mw/m 2 600MW to be radiated in Core and SOL/divertor Q/S=600MW/700m 2 ~1MW/m 2 60MW to divertor plates q=7mw/m 2 RELIABILITY/Robustness is a strong concern!! Transient excursion can easily change by large factor!! 7MW/m 2 is still too large!! 7

8 Divertor Plasma Control (Fluid simula4on) Should be kine-c at SOL!! Imp. force balance ParFcle balance Ion force balance Albedo=0.96 Ion energy balance Electron energy balance Ueda, Kikuchi, et al. NF1992 Bohm diffusion is assumed for SOL particle transport perpendicular to flux surface. 8

9 Do we see significant progress in these 20 years? DEMO : Strong D and impurity puffs at divertor, shallow pellet at SOL Ueda, Kikuchi NF1992 SOL transport : Sophisticated control is required to reduce q~7mw/m 2 even with Bohm diffusion (L-mode) Q=600MW Γ p =2.5x10 23 /s τ E =1.4s τ p =0.5s Fe puff = 0.01Γ p Gas puff 7Γ p Imp. puff 0.01Γ p High Z : sheath accelerafon (important even for He) Stable semi- detach is challenging In reactor : one failure is serious!! Kajita, NF2009 (Top10) W nano structure 9

10 2.2 Heat flux and radia4on heat flux (Fluid simula4on) [1] ConducFon heat flux is ~7MW/m 2 but total heat flux including radiafon heat flux is up to 12MW/m 2. [2] RelaFvely high radiafon heat flux comes from the strong accumulafon of impurity near the divertor plates. Namely, back flow due to thermal force is suppressed by fricfon force. But n Fe /n e is only 0.5-1%. [3] More precise control of impurity injecfon profile is necessary to reduce peak radiafon heat flux. Ueda, Kikuchi, et al. NF Impurity density profile

11 1.2 Old and New Problems 1.2.1: ELM heat flux of H-mode ΔW ~ 20MJ at low collision 2000th - Why it happens : Tokamak MHD is not designed to have soft beta limit : Power e-folding length in H-mode (2 nd Goldston scaling) 2010 th - Why it happens : Fast SOL flow quickly exhaust heat within thin width for low H-mode flux. 11

12 1.2.1 OLD problem : ELM (Edge Localized Mode) Plasma pressure Y. Liang ITER summer school

13 ELM (Edge Localized Mode) If ELM deposit energy in short pulse, tiles will be damaged. No of ELMs Divertor ELM energy density (MJ/m 2 ) Measured ELM energy deposition time: ms A. Loarte NF

14 ELM :RMP(Resonant Magnetic Perturbation) coils Many tokamaks equipped with ELM control coils T. Evans TTF2013 Many issues: " physics : lobes (homoclinic tangle) nonuniform power deposition in toroidal direction" Technologies: neutron damage of coils, localized fast ion loss etc." -> We need to develop method to eliminate ELM without RMP for DEMO. 14

15 1.2.2 Inter-ELM heat flux : Goldston 2 nd scaling R. Goldston NF2012. SOL heat flux e-folding length λ q- SOL 1mm λ q ρ p 5mm R Previous estimate for ITER:5mm Recent estimate for ITER:1mm Figure (Federici, NF2001) Note: L- mode is governed by different physics, empirical scaling 1cm for ITER Div heat flux e- folding length λ q- div is larger by flux expansion ra4o for ajached plasma. 15

16 Inter-ELM heat flux : key physics of Goldston scaling (neo)classical par-cle transport in H- mode Grad /curvature B driv into SOL P SOL is Spitzer thermal conduc4on? <v d >l // = 0.5c s λ Assumed as same order λ <v d > l // <v d > 0.5c s 2 nd Goldston scaling(λ ρ p )! Fast parallel SOL flow reduces λ to 1mm!! A. Chankin NF2007: Fast parallel flow ~ 0.5Cs comes not from fluid simulation, unresolved issue. electron ion 16

17 Experimental result seems in agreement with Goldston scaling C-Mod (B p ~B p ITER ) SOL e-folding length~1mm H-mode particle flux from separatrix ~ neoclassical drift flux. Γ p ELM free H-mode ~0.1 Γ p L- mode B. Lipschultz, FESAC mee4ng July, 2012 Goldston scaling needs more check. 17

18 Grad B away from X: Why SOL flow is so fast as 0.5Cs? Takizuka, NF2009 showed SOL flow is accelerated by both Trapped and Passing ions. Good for impurity control. Bad for SOL heat flux. ion B drij Δ 18

19 How about I-mode (MIT)? I-mode : Grad B away from X-point and need high power L -> I (H) mode High edge T e (low collisionality). L-mode like τ p but at lower edge n e. Note : Reactor needs high SOL n e. [ NSTX Li discharge has high T e and low n e ] Trapped ion orbit Takizuka CPP2010 Whyte NF2010 I-mode geometry has even faster SOL flow -> leads to lower edge density? 19

20 Key questions : 1. Can we increase Γ p H- mode? High recycling at main SOL is prohibitive (wall sputter)! Shallow pellet is still OK. 2. Can we reduce SOL flow speed? Drift across flux surface is key! If we keep fast SOL flow for impurity control, only way is to stay L-mode edge. 3. If not, shall we kill H-mode? L-mode is best but not sufficient Explore improved confinement with L-mode edge (I-mode?). 4. High edge pedestal is good choice? Can we make soft beta limit? (High edge BSC leads to big ELM) If not, shall we reduce edge beta limit for small ELM? 20

21 2. How to design optimum configuration? Power handing the first! Is core the first good design philosophy? 21

22 2.1 : Issues in present reactor design philosophy SSTR1990 (A) : Optimization of Core plasma (B) : Divertor design to match (A) (C) : consistency of (A)& (B) D-shape/H-mode is thought as optimum for CORE. R p 1. D-shape : R div << R p : bad for power handling! 2. H-mode : Large P edge -> Large ELM energy loss! 3. H-mode : Low particle flux! 4. D shape : huge Amp Turn for snow flake. 5. D-shape : inboard blanket design not easy. R div Level of problem : D-shaped > H-mode 22

23 2. Think different! Core the first is not a good design philosophy First priority (A) :Configuration optimization on power handling (B) (1) Core to match (A) (2) Divertor to match (A) (3) Integration to match (A) We have rich knowledge 23

24 First Step : Divertor priority higher than core! Stay hungry, Stay foolish! A choice - negative D - S. Jobes - Make edge pedestal β limit SOFT not by finite n peeling but by Mercier/n= Ballooning! Stay in L-mode edge? Find new transport reduction physics! Ex. Reactor core is more collisionless. Optimization of TEM - Trapped electron precession Negative delta reduce Stiffening 24

25 Make power handling easier by a Large Factor: Case I Negative D + Snowflake Note: Outboard is much easier to install Snowflak R=7m, a=2.7m (A=2.6) Standard D shape : Rx=4.3m, Negative D shape : Rx=9.7m Factor of 2.5 for R div Snow flake at Rx : Factor of Factors : 2.5 x (1.5-3) =

26 Make power handling easier by a Large Factor: CaseII Negative D + Double Null Note: DN in D-shape is difficult for piping to inboard blanket. R=7m, a=2.7m (A=2.6) Standard D shape : Rx=4.3m, Negative D shape : Rx=9.7m Factor of 2.5 for R div DN : Factor of Factors : 2.5 x (1.5-2) =

27 Flux Tube Expansion Divertor (NTT-FTE diveror) Orange: co-current Green : counter-current Blue: flux swing coil Internal FTEcoils ~4MAT each FTE coil pairs produce flux tube expansion by 2.7. Rp=8.5m, ap=2.4m (A~3.5) Ip=15-17MA (ss, hybrid) Bt=6.2T (B max ~13T) For 17MA, β N =3, f GW =0.85, we have P f =3GW. 27

28 Quick summary 1 Can we find a solution for power handling in Tokamak Fusion Reactor? Divertor heat flux : w/o control : ~70MW/m 2 : w control : ~10MW/m 2 New proposal : negative triangularity + Snowflake / Double null or Flux tube ex. Divertor heat flux : w/o control : ~10MW/m 2 : w control : ~2-3MW/m 2 Takizuka JNM

29 3 MHD design 3.1 Small change in d can ELM softer (TCV) 3.2 Negative triangularity tokamak (NTT) can be stable at β N >3. ( Medvedev, Kikuchi, et al., NF2015 ) 3.3 Stability of NTT-FTE reactor 29

30 3.1 MHD stability of negative triangular plasma Negative delta has higher frequency ELM. Pochelon PFR2012 Small tilting of upper triangularity makes difference! 30

31 3.2 Can we achieve reactor relevant β N by NTT? Yes. Magnetic Hill β N >3 is stable" Medvedev, Kikuchi, et al. NF 2015 Can tokamak OK with magnetic hill? Yes, we can get β N >3 #: Mode coupled with Mercier to be internal. LHD magnetic hill - Watanabe NF2005 beta above Mercier - Sakakibara PPCF2008 Res. Interchange m/n=1/1 c.f. Heliotron E has problem with resistive interchange. H-J has magnetic well. 31

32 3.3 How about Single null stability? Axisymmetric mode n The n=0 growth rate with \injy 3.46 a w /a=1.35, (roughly ITER case), β N =3.46" a w /a= > 3.11 A=3.5, κ=1.75, " : 24 /s (~4 times higher δ/δ x = -0.5/-0.9, than for standard ITER Ip=14.88MA, I N =1 configuration)

33 SN NTT-FTE diveror has more flexible shaping Negative/negative Zero/ Negative 33

34 4. Improved confinement with negative delta 4.1 Can we can get improved confinement with L-mode edge? 4.2 TEM stability 4.3 Flow shear suppression 34

35 4.1 improved confinement with L-mode edge? Scenario for improved confinement in NTT - Step 1: Shape and β p optimization to stabilize TEM. Balance between MHD and TEM stabilization is important. - Step 2: Flow shear suppression of turbulence 0-th order force balance equation to produce mean flow shear (If TEM γ is smaller, pressure curvature may be sufficient to produce necessary E x B shear) 35

36 4.2 TEM : Precession de-resonance is key (Kikuchi, Azumi, Rev. Mod. Phys. 2012) Precession drift B.B. Kadomtsev B.B. Kadomtsev, O.P. Pogutse, Rev. Plasma Phys. 1967, NF

37 4.2 TEM : How to stabilize TEM Rosenbluth s maximum J principle (PF1968) Maximum J principle: Growth rate Glasser PF1974" " Elongation (big)" Triangularity" Toroidal shift" are stabilizing" 37

38 4.2 TEM : toroidal drift has strong effect Shafranov shiv will stabilize TEM (Connor NF1983) Shafranov shift can change precession drift 38

39 4.2 TEM : Negative δ and Shafranov shift Negative triangularity will stabilize TEM (Rewoldt PF1982) Negative d can reduce TEM growth rate Charge neutral, Ampere law 39

40 4.2 TEM : Evidence of TEM/ITG transition Wulu Zhong, 2 nd APTWG Tore Supra expl. Dispersion rela4on for TEM/ITG modes in strong ballooning limit. Weiland textbook, 2000 Also, J. Rice, FEC2012 bifurcation of intrinsic rotation TEM/ITG 40

41 4.2 : TCV negative triangularity experiment Camenen NF2007 Negative triangularity produces large Shafranov shift, which changes precession drift of trapped electron. This leads to a change in TEM stability. Large tilting in negative delta! Similar effect like Er? More tilted Less tilted Non- locality will be reduced in Reactor 41

42 4.3 : Flow shear suppression Hahm-Burrell condition: Pressure curvature can produce E x B flow shear 42

43 4.3 : Shaping effect of Residual Zonal Flow (RZF) Xiao- Caqo PoP2006, 2007 Key is to reduce NC polariza4on Belli, Hammeq, Dorland, PoP2008 Radial profile of δ - dδ/dr is key to RZF - NC polarization ~ (Banana width) 2 Negative delta : strong outboard B p -> smaller banana width!! ElongaFon increases RZF NegaFve δ may weakly reduces RZF. Xiao PoP2007 GS2 (1) GS2 (1) Xiao PoP2007 Understanding of RZF in nega-ve triangularity (κ,- δ, Δ) is necessary 43

44 Summary The power system should have reliable power handling but fusion power handling is challenging in divertor. H-mode with D-shaping Optimize Core choice seems enhancing its challenge. Tokamak physics is ready for new innovation. Good knowledge in core physics will make innovation possible. Power handling-driven Tokamak optimization needs good core physics innovation. We proposed Negative D as a candidate of this challenge. This might be foolish idea (S. Jobes) but we need more foolish ideas to solve critical power handling issue. 44