MCZ NCSX & Compact Stellarators in the World Fusion Program M.C. Zarnstorff PPPL Briefing for Prof. R. Fonck Associate Director, Office of Fusion Energy Science 1 May 2007

MCZ Outline What is needed for DEMO? Stellarator solutions Role & goals of NCSX

MCZ What must the World Program develop for DEMO? This requires development of plasmas with: Higher pressure, by at least factor of 1.7 Less external current drive, more bootstrap current; no inductive current No external rotation drive Essentially no disruptions And: high heat flux PFCs, T-breeding cycle, long- lived materials. For tokamaks, many of these needed improvements conflict with each other. ITER (~ 2016) ITER: 500 MW for 10 minutes, gain > 10 DEMO: ~2500 MW, continuous, gain > 25, ~ same size and field.

MCZ Tokamak Advanced Scenarios for Steady-State DEMO Are Very Challenging Most reactor studies aim for f bs ~ 80-99%  requires feedback controlled stability with  >  no-wall and NTM stabilization Results in non-linear coupling between  -heating profile, current profile, transport, and MHD stability  requires feedback controlled pressure/current profile Only feasible solution for divertor power handling appears to be radiating mantle, with P rad ~ 80-90% of loss power  impurity seeding High density operation needed for radiating divertor solution. –EU designs all have n > n Greenwald !! Yet need H H > 1. –Reduces current drive efficiency. See D. Campbell et al. IAEA 2000, Chengdu “Critical Issues for Tokamak Power Plants”

MCZ “Disruptions Will Essentially Have to be Eliminated” D. Campbell et al. IAEA 2006, Chendu “Critical Issues for Tokamak Power Plants” Aries studies: disruptions must be < 1 per year Breeding blanket designs have first-wall thickness of 1.5 – 5 mm insufficient thermal mass to avoid melting, doesn’t stop runaways. How to guarantee no disruptions at high beta in bootstrap sustained configurations?

MCZ Compact Stellarators provide Solutions A steady-state toroidal reactor with Quiescent steady state at high-beta No current drive  intrinsically high Q, can operate ignited Not limited by macroscopic instabilities. No need to control profiles. No need for feedback or rotation to control instabilities, or nearby conducting structure No disruptions  allows thin PFCs, allowing breeding blanket Very high density limit  easier plasma solutions for divertor reduced fast-ion instability drive High power density (similar to ARIES-RS and –AT) Good alpha-particle confinement = Already demonstrated in high-aspect ratio, non-symmetric stellarators Most of these are available in all stellarators (except high beta) Greatly simplifies many aspects of a projected DEMO. Need to demonstrate these properties at low aspect ratio, to project to high power density.

MCZ Stellarator Operating Range is much larger than for Tokamaks  Using equivalent toroidal current that produces same edge iota in Greenwald evaluation.  Limits are not due to MHD instabilities. No disruptions.  High density favorable: – Lower plasma edge temperature, Eases edge design – Reduces energetic particle instability drive

MCZ NCSX Strategy: Combine Advantages of Stellarators and Tokamaks Tokamaks: Compact  cost-effective, project to high power density Importance of flows ( including self-generated) for turbulence stabilization ‘Reversed shear’ to reduce turbulence, increase stability, suppress islands Stellarators: Externally-generated helical fields –Plasma current not required. No current drive. Steady-state easy. –Robust stability. Generally, disruption-free. Numerical optimization of 3D field (shape) to obtain desired physics properties, including –Increased stability, good flux surfaces at high-beta –Quasi-toroidal symmetry: If |B| is symmetric in flux coordinates, orbits are like tokamak  neoclassical transport very similar to tokamaks, undamped rotation Goal: Steady-state high- , good confinement without disruptions

MCZ NCSX Designed to Integrate Attractive Properties At Low Aspect Ratio & Address Needs 3 periods, R/  a  =4.4,  ~1.8,  ~1 Quasi-axisymmetric: transport similar to tokamaks ripple thermal transport insignificant. Passively stable at  =4.1% to kink, ballooning, vertical, Mercier, neoclassical-tearing modes, … (steady-state tokamak limit ~ 2.7% without feedback) Stable for  > 6.5% by adjusting coil currents Passive disruption stability: equilibrium maintained even with total loss of  or I P Flexible configuration: 9 independent coil currents by adjusting currents can control stability, transport, shape: iota, shear The US is the World Leader in Compact Stellarator Design

MCZ ARIES-CS: a Competitive, Attractive Reactor Reference parameters for baseline: NCSX-like config.  R  = 7.75 m  a  = 1.72 m  n  = 4.0 x m –3  T  = 6.6 keV  B  axis = 5.7 T  = 6.4% H(ISS95) = 2.0 H(ISS04) = 1.1 I plasma = 3.5 MA (bootstrap) P(fusion) = GW P(electric) = 1 GW Based on NCSX design Aries--I-RS-CS-AT-CS BlanketLiPb/FSLiPb/SiC COE(92)

MCZ NCSX Research Mission Acquire the physics data needed to assess the attractiveness of compact stellarators; advance understanding of 3D fusion science. Understand… Pressure limits and limiting mechanisms in a low-A optimized stellarator Effect of 3D magnetic fields on disruptions. Operating limits. Reduction of and anomalous neoclassical transport by quasi-axisymmetric design. Confinement scaling; reduction of turbulent transport by flow shear control. Equilibrium islands and tearing-mode stabilization by design of magnetic shear. Compatibility between power and particle exhaust methods and good core performance in a compact stellarator. Energetic-ion stability and confinement in compact stellarators Demonstrate… Conditions for high , disruption-free operation High pressure, good confinement, compatible with steady state Can we achieve the promise of Aries-CS? Can the design be simplified through improved understanding?

MCZ The World’s Stellarator Programs Focused on Superconducting, Steady State Large Helical Device (Japan) Enhanced confinement, high pressure; A = 6-7, R=3.9 m, B=3T Wendelstein 7-X (Germany) (2012) Non-symmetric optimized design: A = 11, R=5.4 m, B=3T Significant technology programs on 3D superconducting coils LHD has achieved 54-minute pulses. Large aspect ratio. Neither optimized for symmetry  flows strongly damped Both are very interested in NCSX, collaborating & contributing hardware ( ECH from Germany; HIBP from Japan under discussion )

MCZ Confinement Degrades with Ripple  eff NCSX Has Lowest New global confinement scaling study for stellarators (ISS04v3) found strong dependence on ripple magnitude (  eff ). NCSX designed for the lowest ripple of all configurations. HSX has demonstrated advantages of quasi-symmetry: increased confinement and decreased flow damping ~  eff –0.4 ? 1/ transport ~  eff 3/2 NCSX ARIES-CS

MCZ

MCZ ‘Reversed Shear’ Key to Enhanced Stability in NCSX Quasi-axisymmetry  tokamak like bootstrap current ~3/4 of transform (poloidal-B) from external coils  externally controllable Rotational transform rising to edge key for stabilizing trapped particle and neoclassical tearing instabilities Explored locally on tokamaks, but cannot be achieved across whole plasma using current Safety facto)r (q) Radial Coordinate 2

MCZ Turbulence Growth Decreases for Higher  p Similar to Reversed Shear Tokamak  Designed for ‘reversed shear’ to help stabilize turbulent transport  Linear ion-scale turbulence growth rates calculated by FULL-code:  Electron-drive stabilized by reversed shear  Ion-drive strongly reduced with   Similar to reversed shear tokamak  Very low effective helical ripple gives low flow-damping allows efficient flow-shear stabilization, control of E r. Allows strong zonal flows (Mynick). - G. Rewoldt

MCZ LHD and W7AS both achieved high-  without being designed for it! LHD recently achieved  =5% In both cases, well above theoretical linear stability limit < 2%. Quiescent. MHD activity not limiting. No disruptions observed. Sustained without CD. Germany Japan

MCZ S. Sakaibara, Y. Suzuki M.C. Zarnstorff, A. Reiman NCSX is designed to keep good flux surfaces and linear stability at high   -limit Correlates with Destruction of Outer Flux Surfaces

MCZ NCSX Designed to Produce Good Flux Surfaces at High-  Poincare: PIES, free boundary without pressure flattening < 3% flux loss, including effects of reversed shear and || vs.  transport. Explicit numerical design to eliminate resonant field perturbations ‘Reversed shear’ configuration  pressure-driven plasma currents heal equilibrium islands (not included in figure) Computation boundary

MCZ Initial Non-Linear Kinetic Calculations Indicate Possible Higher Pressure-Limit for NCSX Magnetic Flux Surfaces ExB Flow Surfaces NCSX Preliminary M3D calculations. Fixed boundary. Finite gyro-radius and self- generated flows stabilize equilibrium Does not include neoclassical effects yet. Should further increase stabilization. What will the pressure limit be for NCSX?  = 7%

MCZ Intrinsic Divertors in Bean-tips divertor vacuum vessel Divertors already operated successfully in LHD and W7-AS. Controlled exhaust and impurities. Strong flux-expansion always observed in NCSX bean-shaped cross-section. Allows isolation of PFC interaction. Similar to expanded boundary shaped-tokamak configurations SOL connection length can be ~100m. Long enough to ensure low temperature divertor plasma. pumps MFBE field-line tracing

MCZ Modular Coils + Toroidal Solenoid + Poloidal Coils for shaping control & flexibility Useful for testing understanding of 3D effects in theory & determining role of iota-profile E.G., can use coils to vary –effective ripple by factor > 10. –Avg. magnetic shear by factor > 5 –Edge rotational transform by factor of 2 Reduce kink-instability threshold down to  1% by modifying plasma shape –either at fixed shear or fixed edge-iota ! These types of experiments will be key for developing and validating our understanding NCSX Coils Designed for Flexibility Shear Rotational Transform

MCZ Conclusions Compact stellarators provide solutions for DEMO –ARIES-CS competitive with best AT designs –No disruptions, no current-drive, increased pressure, no profile feedback. –Relatively small step from achieved stellarator characteristics NCSX is an exciting opportunity for unique fusion-science research. –High- , compatible with steady state, good flux surfaces, and stability –Sustainment without external current drive –Stability without rotation drive or active feedback on instabilities –No disruptions –Tokamak-like transport using quasi-axisymmetry: low rotation damping, good alpha confinement. NCSX and compact stellarators provide an opportunity for US Leadership and a route to a more practical DEMO.

MCZ Supplemental

MCZ Disruptions are a Big Issue Must design to protect facility & environment in case of disruption. ITER: PFCs use 10mm of Be or W, backed by 22mm of Cu to have sufficient thermal mass to prevent bulk melting -- incompatible with breeding blankets -- Calculations of disruption PFC erosion give 0.25 mm of W for a VDE-down onto baffle 0.15 mm of Be for a VDE-up 0.10 mm of Be for a major disruption (last weeks WG 1 meeting; Sugihara, NF 47 (2007) 337) Electromagnetic loads are <10% from allowable, strongly dependent on width of halo currents Even mitigated disruptions are calculated to melt the PFC surfaces, may produce cracking Disruptions in advanced scenarios (reversed-shear or above no-wall limit) are often prompt, with shortest quench times. (draft ITER Physics Basis, Ch. 3) -- may not be possible to predict or mitigate -- ITER (~ 2016)

MCZ Wide Range of  and * Accessible B = 1.2 T, 3MW *  =2.7%, *I =0.25 with H ISS95 =2.9; H ISS04 =1.5 H ITER-97P =0.8 *  =2.7%, *I =2.5 with H ISS95 =2.0; H ISS04 =1.0 *  =1.4%, collisional with H ISS95 =1.0, ; H ISS04 =0.5 sufficient to test stability theory Contours of H ISS95, H ITER-97P, and min *i LHD and W7-AS have achieved H ISS95 ~ 2.5 PBX-M obtained  = 6.8% with H ITER-97P = 1.7 and H ISS95 ~ 3.9 * * * n e (10 19 m -3 ) (%)

MCZ High-  low * Plasmas Accessible (6MW) B = 1.2 T, 6MW  =4%, *I =0.25 requires H ISS95 =2.9, H ISS04 =1.5 H ITER-97P =0.9  =4% at Sudo-density H ISS95 =1.8, H ISS04 =0.9  H ISS95 =1.0 gives  =2.2% at high collisionality Contours of H ISS95, H ITER-97P, and min *i LHD and W7-AS have achieved H ISS95 ~ 2.5 PBX-M obtained  = 6.8% with H ITER-97P = 1.7 and H ISS95 ~ 3.9 (%) n e (10 19 m -3 )

MCZ How to Achieve Orbit Confinement in 3D?? Quasi-symmetry 3D shape of standard stellarators No symmetry  no conserved canonical momenta orbits can have resonant perturbations, become stochastic  lost B is bumpy every direction  rotation is strongly damped  ‘Quasi-symmetry’  (Boozer,1983) Orbits & collisional transport depends on variation of |B| within flux surface, not the vector components of B !  Hamiltonian for particle drift-motion only depends on |B| in flux coordinates  (Nührenberg) If |B| is symmetric in flux coordinates, get confined orbits like tokamak  neoclassical transport very similar to tokamaks (theoretically), undamped rotation Recently tested in HSX, at Univ. Wisconsin –quasi-Helical-Symmetry reduces electron transport, flow damping –Too small to study high pressure or ion transport

MCZ NCSX Design: Detailed Plasma Modeling & Optimization Vary Coil Shapes ~ 200–400 free parameters Coil Characteristics 3D Equilibrium Calc. ( VMEC ) Macroscopic Stability High- & low-n Orbit Confinement Flux Surface Quality Ripple Transport... Adjust Coil Shapes & Currents ( Levenberg-Marquardt, Differential Evolution, Genetic ) Direct design of coil shapes to achieve desired physics properties Only possible using high-capacity advanced computing and advances in theoretical understanding and modeling

MCZ  ≈ 3.4 % : Quiescent, Quasi-stationary  B = 0.9 T, iota vac ≈ 0.5  Almost quiescent high-  phase, MHD-activity in early medium-  phase  In general,  not limited by any detected MHD-activity.  I P = 0, but there can be local currents  Peak  ~8%  Similar plasmas with B = 0.9 – 1.1 T, either NBI-alone, or combined NBI + OXB ECH.  Much higher than predicted linear stability  limit ~ 2% 54022

MCZ For compact, quasi-symmetric, sustainable high-beta configurations: 1.Can beta ~6.4% be achieved and sustained at good confinement? What is the maximum useful beta? 2.Can low alpha loss be achieved? Can alpha loss due to MHD instabilities be mitigated by operation at high density? 3.Develop a workable divertor design with moderate size and power peaking, that controls impurities and enables ash pumping. 4.Demonstrate regimes of minimal power excursions onto the first wall (e.g. due to disruptions and ELMs). 5.Under what conditions can acceptable plasma purity and low ash accumulation be achieved? 6.Is the energy confinement at least 2 times ISS95 scaling? How does it extrapolate to larger size? 7.Characterize other operational limits (density, controllable core radiation fraction) 8.How does the density and pressure profile shape depend on configuration and plasma parameters? 9.Can the coil designs be simplified? Can physics requirements be relaxed, by a.Reduction of external transform b.Elimination of stability from optimization c.Reducing flux-surface quality requirements d.Increased helical ripple 10.What plasma control elements and diagnostics are required? NCSX Research Plan is focused on addressing these questions and developing their scientific basis. ARIES-CS Physics R&D Needs