Optimization of Compact Stellarator Configuration as Fusion Devices Report on ARIES Research

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1 Optimization of Compact Stellarator Configuration as Fusion Devices Report on ARIES Research Farrokh Najmabadi and the ARIES Team UC San Diego OFES Briefing November 16, 2005 Germantown Electronic copy: ARIES Web Site:

2 UC San Diego Boeing GA GIT INEL MIT ORNL PPPL RPI U.W. FKZ Collaborations For ARIES Publications, see: For ARIES Publications, see:

3 ARIES-Compact Stellarator Program Has Three Phases FY03/FY04: Exploration of of Plasma/coil Configuration and and Engineering Options Develop physics physics requirements and and modules (power (power balance, balance, stability, α confinement, divertor, etc.) etc.) Develop engineering requirements and and constraints Explore Explore attractive coil coil topologies. Present status FY06: FY06: Detailed system system design design and and optimization FY04/FY05: Exploration of of Configuration Design Design Space Space Physics: Physics: β, β, A, A, number number of of periods, periods, rotational transform, sheer, sheer, etc. etc Engineering: configuration optimization, management of of space space between between plasma plasma and and coils, coils, etc. etc Trade-off Studies Studies (Systems Code) Code) Choose Choose one one configuration for for detailed detailed design. design.

4 Goal: Stellarator Power Plants Similar in Size to Tokamak Power Plants Multipolar external external field field -> -> coils coils close close to to the the plasma plasma First First wall/blanket/shield set set a a minimum plasma/coil distance distance (~1.5-2m) A minimum minor minor radius radius Large Large aspect aspect ratio ratio leads leads to to large large size. size. Plasma Aspect Ratio <R>/<a> Need a factor of 2-3 reduction Need a factor of 2-3 reduction ARIES AT Tokamak Reactors SPPS HSR-4 FFHR-1 MHR-S HSR-5 Stellarator Reactors Compact Stellarator Reactors ARIES RS Circle area ~ plasma area Average Major Radius <R> (m) Approach: Physics: Reduce Reduce aspect aspect ratio ratio while while maintaining good good stellarator properties. Engineering: Reduce Reduce the the required required minimum coil-plasma distance.

5 We have focused on Quasi-Axisymmetric stellarators that have tokamak transport and stellarator stability In In 3-D 3-D magnetic field field topology, particle drift drift trajectories depend only only on on the the strength of of the the magnetic field field not not on on the the shape of of the the magnetic flux flux surfaces. QA QA stellarators have have tokamak-like field field topology. Stellarators with with externally supplied poloidal flux flux have have shown resilience to to plasma disruption and and exceeded stability limits predicted by by linear theories. QA QA can can be be achieved at at lower aspect ratios with with smaller number of offield field periods. A more more compact device (R<10 m), m), Bootstrap can can be be used used to to our our advantage to to supplement rotational transform, Shown to to have have favorable MHD stability at at high high β. β.

6 Typical Plasma Configuration Optimization Criteria Maximum residues of of nonaxisymmetry in in magnetic spectrum. neo-classical neo-classical transport transport << << anomalous anomalous non- transport: transport: overall overall allowable allowable noise noise content content < ~2%. ~2%. effective effective ripple ripple in in 1/ν 1/νtransport, εε eff ~1% eff < ~1% ripple ripple transport transport and and energetic energetic particle particle loss loss α energy energy loss loss < ~10% ~10% Equilibrium and and equilibrium β limits limits Shafranov Shafranovshift shift < β > A < a> 2 2κι < 1/2 large large islands islands associated associated with with low low order order rational rational surfaces surfaces flux flux loss loss due due to to all all isolated isolated islands islands < 5% 5% overlapping overlapping of of islands islands due due to to high high shears shears associated associated with with the the bootstrap bootstrap current current limit limit dι/ds dι/ds Stability limits limits (linear, (linear, ideal ideal MHD) MHD) vertical vertical modes modes 2 ι ext κ κ / ι κ interchange interchange stability: stability: V ~2-4%. V ~2-4%. LHD, LHD, CHS CHS stable stable while while having having a a hill. hill. ballooning ballooning modes: modes: stable stable to to infinite-n infinite-n modes modes LHD LHD exceeds exceeds infinite-n infinite-n results. results. High-n High-n calculation calculation typically typically gives gives higher higher ββ limits. limits. kink kink modes: modes: stable stable to to n=1 n=1 and and 22 modes modes without without a a conducting conducting wall wall W7AS W7AS results results showed showed mode mode (2,1) (2,1) saturation saturation and and plasma plasma remained remained quiescent. quiescent. tearing tearing modes: modes: dι/ds dι/ds > 00 Each Each criteria criteria is is assigned a a threshold and and a a weight weightin in the the optimization process. process.

7 Stellarator Operating Limits Differ from Tokamaks Stellarators operate operate at at much much higher higher density density than than tokamaks Limit Limit not not due due to to MHD MHD instabilities. Density Density limited limited by by radiative recombination High-β High-βis is reached reached with with high high density density (favorable density density scaling scaling in in W7-AS) High High density density favorable for for burning burning plasma/power plant: plant: Reduces edge edge temperature, eases eases divertor divertor solution solution Reduces α pressure pressure and and reduces reduces α- α- particle particle instability drive drive Greenwald density evaluated using equivalent toroidal current that produces experimental edge iota

8 Stellarator β May Not Limited by Linear Instabilities β β > % for for > ττ E E (W7AS) β β > % for for > ττ E (LHD) E (LHD) Peak Peak β β Average flat-top flat-top β β very very stationary plasmas plasmas No No Disruptions Duration and and β not not limited limited by by onset onset of of observable MHD MHD Much Much higher higher than than predicted β limit limit of of ~ 2% 2% (from (from linear linear stability) 2/1 2/1 mode mode ovserved, but but saturates. No No need need for for feedback mode mode stabilization, internal internal coils, coils, nearby nearby conducting structures. β-limit β-limit may may be be due due to to equilibrium limits. limits. <β> (%) <β> peak <β> flat-top avg τ flat-top / τ E

9 Physics Optimization Approach NCSX scale-up Coils 1) Increase plasma-coil separation 2) Simpler coils High leverage in sizing. Physics 1) Confinement of α particle 2) Integrity of equilibrium flux surfaces Critical to first wall & divertor. Reduce consideration of MHD stability in light of W7AS and LHD results New classes of QA configurations MHH2 1) Develop very low aspect ratio geometry 2) Detailed coil design optimization How compact a compact stellarator power plant can be? SNS 1) Nearly flat rotational transforms 2) Excellent flux surface quality How good and robust the flux surfaces one can design?

10 Optimization of NCSX-Like Configurations: Increasing Plasma-Coil Separation A A series of of coil coil design with with A c min 6.8 to 5.7 c =<R>/ min ranging 6.8 to 5.7 produced. Large increases in in B max only for c 6. max only for A c < 6. α α energy loss loss is is large large ~18%.. LI383 A c =5.9 For <R> = 8.25m: min (c-p)=1.4 m min (c-c)=0.83 m I max =16.4

11 Optimization of NCSX-Like Configurations: Improving α Confinement & Flux Surface Quality A bias bias is is introduced in in the the magnetic spectrum in in favor of of B(0,1) A A substantial reduction in in α loss loss (to (to ~ 3.4%) is is achieved. LI383 N3ARE N3ARE Frequency *4096 Energy (kev) Energy (kev) The external kinks and and infinite-n ballooning modes are are marginally stable at at 4% 4% βwith no no nearby conducting wall. wall. Rotational transform is is similar to to NCSX, so so the the same same quality of of equilibrium flux flux surface is is expected.

12 Optimization of NCSX-Like Configurations: Improving α Confinement & Flux Surface Quality The The external transform is is increased to to remove m=6 m=6 rational surface and and move m=5 m=5 surface to to the the core core Equilibrium calculated by β. KQ26Q May be be unstable to to free-boundary modes but but could be be made more more stable by by further flux flux surface shaping

13 Two New Classes of QA Configurations II. II. MHH2 Low plasma aspect ratio ratio (A (A p 2.5) in field p ~ 2.5) in 2 field period. Excellent QA, QA, low low effective ripple (<0.8%), low low α energy loss loss ( ( 5%).. III. III. SNS SNS A p 6.0 in field period. QA, low (< loss 8%. p ~ 6.0 in 3 field period. Good QA, low effective ripple (< 0.4%), α loss 8%. Low shear rotational transform at at high high β, β, avoiding low low order resonances.

14 α loss is still a concern Issues: Issues: High High heat heat flux flux (added (added to to the the heat heat load load on on divertor divertor and and first first wall) wall) Material Material loss loss due due to to accumulation accumulation of of He He atoms atoms in in the the armor armor (e.g., (e.g., Exfoliation Exfoliation of of µm µm thick thick layers layers by by MeV MeVα s): Experiment: Experiment: He He Flux Flux of of 22 x x /m /m 2 s 2 s led led to to exfoliation exfoliation of of 3µm 3µm W layer layer once once per per hour hour (mono-energetic (mono-energetic He He beam, beam, cold cold sample). sample). For For GW GW of of fusion fusion power, power, 5% 5% αloss, and and α s α sstriking striking 5% 5% of of first first wall wall area, area, ion ion flux flux is is x x /m /m 2 s). 2 s). Exact Exact value value depend depend on on αenergy spectrum, spectrum, armor armor temperature, temperature, and and activation activation energy energy for for defects defects and and can can vary vary by by many many orders orders of of magnitude magnitude (experiments (experiments and and modeling modeling needed). needed). Footprints of escaping α on LCMS for N3ARE. Heat load and armor erosion maybe localized and high

15 Minimum Coil-plasma Stand-off Can Be Reduced By Using Shield-Only Zones

16 The Radial Build Transition coil structure For For NCSX-type configurations, coils coils are are far far from from the the plasma except for for ~5% ~5% of of the the wall wall area, area, which allows a shield-only zone zone in in that that area area and and hence a smaller value for for <R <R axis axis > MHH2 coil coil configurations do do not not allow this this

17 Resulting power plants have similar size as Advanced Tokamak designs Trade-off between good good stellarator properties (steady-state, no no disruption,, no no feedback stabilization) and and complexity of of components. Complex interaction of of Physics/Engineering constraints.

18 Desirable plasma configuration should be produced by practical coils with low complexity Complex 3-D 3-D geometry introduces severe severe engineering constraints: Distance between between plasma plasma and and coil coil Maximum coil coil bend bend radius radius Coil Coil support support Assembly and and maintenance

19 Coil Complexity Impacts the Choice of Superconducting Material Strains Strains required required during during winding winding process process is is too too large. large. NbTi-like (at (at 4K) 4K) B < ~7-8 ~7-8 T NbTi-like (at (at 2K) 2K) B < 9 T, T, problem problem with with temperature margin margin Nb Nb 3 Sn or MgB 2 16 T, Wind React: 3 Sn or MgB 2 B < 16 T, Wind & React: Need Need to to maintain structural integrity during during heat heat treatment (700 (700 o o C for for a a few few hundred hundred hours) hours) Need Need inorganic insulators Inorganic insulation, assembled with with magnet magnet prior prior to to winding winding and and thus thus capable capable to to withstand the the Nb Nb 3 Sn heat process. 3 Sn heat treatment process. Two Two groups groups (one (one in in the the US, US, the the other other one one in in Europe) Europe) have have developed glasstaptape that that can can withstand the the process glass- process A. Puigsegur et al., Development Of An Innovative Insulation For Nb3Sn Wind And React Coils

20 Coil Complexity Dictates Choice of Magnet Support Structure It It appears that that the the out-of-plane force force are are best best supported by by a continuous structure with with superconductor coils coils wound into into grooves Net Net force force balance between field field periods Winding is is internal internal to to the the structure, projection on on the the outer outer surface surface is is shown. shown.

21 Because of Complex Shape of Components Assembly and Maintenance Is a Key Issue

22 Field-Period Assembly: Components are replaced from the ends of field-period Takes Takes advantage of of net net force force balance balance in in a a field field period period Drawbacks: Complex shield shield (lifetime components) geometry. Very Very complex initial initial assembly (of (of lifetime lifetime components) Complex warm/cold interfaces (magnet (magnet structure) and/or and/or magnet magnet should should be be warmed warmed up up during during maintenance. Life-time components (shield) (shield) should should be be shaped shaped so so that that replacement components can can be be withdrawn. CAD CAD exercises are are performed to to optimize shield shield configuration.

23 Port Assembly: Components are replaced Through Three Ports Modules removed through through three three ports ports using using an an articulated boom. boom. Drawbacks: Coolant Coolant manifolds increases plasma-coil distance. Very Very complex manifolds and and joints joints Large Large number number of of connect/disconnects

24 Blanket Concepts are Optimized for Stellarator Geometry Dual Dual coolant with with a self-cooled PbLi PbLizone and and He-cooled RAFS structure Originally developed for for ARIES-ST, further further developed by by EU EU (FZK), (FZK), now now is is considered as as US US ITER ITER test test module module SiC SiC insulator lining lining PbLi PbLichannel for for thermal thermal and and electrical insulation allows allows a a LiPb LiPboutlet temperature higher higher than than RAFS RAFS maximum temperature Self-cooled PbLi PbLiwith SiC SiC composite structure (a (a al al ARIES-AT) Higher-risk high-payoff option

25 Divertor Design is Underway Several Several codes codes (VMEC, MFBE, MFBE, GOURDON, and and GEOM) GEOM) are are used used to to estimate the the heat/particle flux flux on on the the divertor divertor plate. plate. Because Because of of 3-D 3-D nature nature of of magnetic topology, location location & shaping shaping of of divertor divertor plates plates require require considerable iterative iterative analysis. W alloy inner cartridge W armor W alloy outer tube Divertor module module is is based based on on W Cap Cap design design (FZK) (FZK) extended to to mid-size (~ (~ cm) cm) with with a a capability of of MW/m MW/m 2 2

26 Summary New New configurations have have been been developed, others refined and and improved, all all aimed at at low low plasma aspect ratios (A (A 6), hence compact size: size: Both Both 2 and and 3 field field periods possible. Progress has has been been made to to reduce loss loss of of α particles to to 5%; this this may may be be still still higher than than desirable. Resulting power plants have have similar size size as as Advanced Tokamak designs. Modular coils coils were were designed to to examine the the geometric complexity and and the the constraints of of the the maximum allowable field, field, desirable coil-plasma spacing and and coil-coil spacing, and and other other coil coil parameters. Assembly and and maintenance is is a key key issue issue in in configuration optimization. In In the the integrated design phase, we we will will quantify the the trade-off between good good stellarator properties (steady-state, no no disruption, no no feedback stabilization) and and complexity of of components.