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Steady state tokamak research ( Power and particle handling – Is H-mode relevant for fusion reactor?) M. Kikuchi Supreme Researcher, JAEA Chairman, Nuclear.

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Presentation on theme: "Steady state tokamak research ( Power and particle handling – Is H-mode relevant for fusion reactor?) M. Kikuchi Supreme Researcher, JAEA Chairman, Nuclear."— Presentation transcript:

1 Steady state tokamak research ( Power and particle handling – Is H-mode relevant for fusion reactor?) M. Kikuchi Supreme Researcher, JAEA Chairman, Nuclear Fusion Board of Editors Guest Professor, ILE Osaka University Visiting Professor, Fudan University, SWIP Guest Lecturer, the University of Tokyo Acknowledgement: A. Fujisawa for turbulence & measurement L. Villard, A. Fasoli and TCV team R. Goldston for SOL heat flow scaling J. Rice, B. Lipschultz for C-Mod, I-mode H. Sugama for NC polarization Wulu Zhong/X. Duan for ITG/TEM work Pat Diamond for discussion (WCI symposium) Lecture 3 at ASIPP, May 15, 2013

2 Motivation of this talk 1.“Tokamak” is a most promising concept with its excellent energy confinement. 1.Tokamak with D-shaped, H-mode is optimized for core confinement. 3. Steady state operation needs more work (see my Reviews of Modern Physics (2012).

3 Motivation of this talk 1.Recent papers by Goldston (NF2012) and Eich(PRL2011) casted important question on reactor power handling in H-mode. Prediction for ITER heat flux 1/e length q-SOL =5mm -> 1mm. 2. ITER may be able to manage power handling for lower P f ~0.5GW and short pulse t duration ~400s by temporary measures such as RMP, pellet pace making, etc. 3. But DEMO/Commercial requires P f ~3GW & t duration ~10Months. This may require fundamental change in design philosophy for tokamak reactor configuration. “Optimize CORE” -> “Optimize power handling”.

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

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

7 Divertor Plasma Control (Fluid simulation) Albedo=0.96 Particle balance Ion force balance Ion energy balance Electron energy balance Imp. force balance Ueda, Kikuchi, et al. NF1992 Bohm diffusion is assumed for SOL particle transport perpendicular to flux surface. Should be kinetic at SOL !!

8 Where is question on power handling? Figure (Federici, NF2001) Previous estimate for ITER:5mm Recent estimate for ITER:1mm Div heat flux e-folding length q-div is larger by flux expansion ratio for attached plasma. R. Goldston NF2012. H-mode SOL Note: L-mode is governed by different physics, empirical scaling 1cm for ITER SOL heat flux e-folding length q-SOL R 1mm 5mm q  p

9 What is key physics of Goldston scaling? ionelectron (neo)classical particle transport in H-mode Assumed as same order l // 0.5c s ✪ Grad /curvature B drift into SOL ✪ Parallel flow connect top and bottom ✪ P SOL is Spitzer thermal conduction 2 nd Goldston scaling(  p ) Fast parallel SOL flow reduces to 1mm!! A. Chankin NF2007: Fast parallel flow ~ 0.5Cs comes not from fluid simulation, unresolved issue.

10 B. Lipschultz, FESAC meeting July, 2012 “ Goldston scaling needs more check.” C-Mod (B p ~B p ITER ) SOL e-folding length~1mm Key evidences : 1.H-mode particle flux from separatrix ~ neoclassical drift flux. 1.Particle flux  p ELM free H-mode ~ 0.1  p L-mode is too low and, Required flux multiplication factor G becomes larger. T div ~ q //div / (G  p / n ) 3.Scale length difference n >> q especially in H-mode 4. ELM to enhance  p : ELM must be minute. Controllability of ELM  p << L-mode Experimental result seems in agreement with Goldston scaling

11 Why SOL flow is so fast as 0.5Cs ? Takizuka, NF2009 showed PARASOL PIC simulation reproduces correct SOL flow pattern and fast SOL flow but not Er effect. Trapped & Circulating ion excursion across the separatrix comparably kick parallel ion flow to be 0.5Cs like a NC parallel viscous force!! Takizuka, CPP2010 (PET12) - It is ion convective flux !! -

12 Key questions : 1.Can we increase  p H-mode ? High recycling at main SOL is prohibitive! 2. Can we reduce SOL flow speed? Drift across flux surface is key! 3.If not, shall we kill H-mode? L-mode is best but not sufficient I-mode as an alternative path? 4. High edge pedestal is good choice? Shall we reduce edge beta limit for small ELM? Key questions : 1.Can we increase  p H-mode ? High recycling at main SOL is prohibitive! 2. Can we reduce SOL flow speed? Drift across flux surface is key! 3.If not, shall we kill H-mode? L-mode is best but not sufficient I-mode as an alternative path? 4. High edge pedestal is good choice? Shall we reduce edge beta limit for small ELM?

13 Modify H-mode to more high recycling? [1] Wall saturation is natural consequence of steady state tokamak reactor. [2] Ti at mid-plane SOL is order of eV, strong gas puff at mid-plane produces energetic neutrals to erode wall a few cm/year. [3] DEGAS simulation in typical JT-60U condition showing non-negligible population of fast neutrals ( eV). [4] Therefore control of neutral around main first wall is important. Kikuchi, FED2006 Gas puffing at main chamber is prohibitive!!

14 Issues in present reactor design philosophy (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. 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. D-shape/H-mode is thought as optimum for CORE. 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. SSTR1990 RpRp R div Level of problem : D-shaped > H-mode

15 I-mode (MIT) with peaked n e may be better, but -- 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  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?

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

17 First Step : Divertor priority higher than core! Stay foolish ! A choice - negative D Make edge pedestal  limit low! Stay in L-mode edge or I-mode? Find new transport reduction physics! Ex. Reactor core is more collisionless. Optimization of TEM - Trapped electron precession Negative D reduce TEM growth. - S. Jobes -

18 Make power handling easier by an order of magnitude R=7m, a=2.7m (A=2.6) Standard D shape : Rx=4.3m Inverted D shape : Rx=9.7m Factor of 2.5 for R div Negative D makes DN possible Factor of (care on up-down asymmetry, controllability) Snow flake at Rx : Factor of 2-3 Factors : 2.5 x 2 x 2 =10 !!! 4.3m Note: - DN in D-shape is difficult for piping to inboard blanket. - Snowflake needs internal PF coil to reduce AT. - Outboard is much easier to install internal PF. Field becomes stiff by near-by PF coils NbTi is possible at low field. 9.7m

19 MHD stability of negative triangular plasma Negative delta has higher frequency ELM. Strongly shaped negative delta has higher edge pressure limit at low J // / due to large shear. Pochelon PFR2012 Courtesy : TCV team

20 Structure of SOL flow in negative D High field side: There is no trapped particles across Separatrix. -> Absence of parallel acceleration mechanism -> Absence of subsonic flow? High field side: There is no trapped particles across Separatrix. -> Absence of parallel acceleration mechanism -> Absence of subsonic flow? Low field side: SOL is almost vertical -> No NC drift across separatrix. -> No change in pressure anisotropy -> Do we see parallel viscous force? Larger local pitch -> shorter connection L Near X-point -> lower local pitch by snow flake Low field side: SOL is almost vertical -> No NC drift across separatrix. -> No change in pressure anisotropy -> Do we see parallel viscous force? Larger local pitch -> shorter connection L Near X-point -> lower local pitch by snow flake I p, B t

21 Banana orbit loss in negative D Confined Banana : Larger than banana width from separatrix, trapped ions will be confined. Confined Banana : Larger than banana width from separatrix, trapped ions will be confined. Lost Banana: Near the separatrix, we have lost banana orbit. -> This may induce Er > standard D. -> Effective RWM stabilization. -> Nullify parallel flow acceleration in low field SOL. Lost Banana: Near the separatrix, we have lost banana orbit. -> This may induce Er > standard D. -> Effective RWM stabilization. -> Nullify parallel flow acceleration in low field SOL. I p, B t

22 2 nd Step : Consistent core plasma! There are two paradigm to suppress turbulent transport 1.Flow shear/zonal flow suppression 2.De-resonance of trapped particle precession with TEM Operationally, we have 3 core improved regimes (See my RMP paper) 1.Weak positive shear (High  p mode, optimized shear, improved H, etc) 2. Negative shear (NS, RS, NCS, etc) 3. Current Hole See Fujita NF review paper.

23 B.B. Kadomtsev, NF 1971 Connor, NF 1983 Negative  and Shafranov shift Precession drift Good for high  p scenario since Shafranov shift increases with  p Shafranov shift can change precession drift Negative  can reduce TEM growth rate G. Rewoldt, PF 1982

24 Dispersion relation for TEM/ITG modes in strong ballooning limit. Weiland textbook, 2000 Wulu Zhong, 2 nd APTWG Tore Supra expl. Increasing experimental evidence of TEM/ITG transition Also, J. Rice, FEC2012 bifurcation of intrinsic rotation TEM/ITG

25 Shaping effect of Residual Zonal Flow (RZF) Xiao-Catto PoP2006, 2007 Belli, Hammett, Dorland, PoP2008 Elongation increases RZF Negative  may weakly reduces RZF. Radial profile of  - d  /dr is key to RZF - Understanding of RZF in negative triangularity ( ,- ,  is necessary Xiao PoP2007 (1) Key is to reduce NC polarization (1) GS2 NC polarization ~ (Banana width) 2 Negative delta : strong outboard B p -> smaller banana width!!

26 Kikuchi NF1990, PPCF1993 Ozeki IAEA1992 FujitaPRL2001,05 OzekiEPS2011 FujitaNF2011 Wall stab. q(0) up Reduce dp/dr at q min Core improved confinements WS regimeNS regime CH regime

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

28 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.

29 Prof. P.H. Rebut : Best Scientist in engineering and physics He is in favor of Fusion-Fission Hybrid. I asked him why? P.H. Rebut : There is no solution for power handing in pure fusion, right now. Stay low fusion power. We have to boost fusion energy to have net energy. Fission is most effective to boost. His word is important from engineering point of view on pure fusion. We probably need order of magnitude change to solve this issue.


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