D ivertor issues on next step devices, and validating the Super-X divertor as a promising solution

1. Divertor issues on next step devices, and validating the Super-X divertor as a promising solution M. Kotschenreuther, S. Mahajan, P. Valanju, Institute for Fusion Studies, University of Texas at Austin J. Canik, R. Maingi Oak Ridge National Laboratory B. LaBombard MIT Garafalo General Atomic Work supported by US DOE-OFES Grant No. DE-FG02-04ER54742 and DOE-ICC Grant No. DE-FG02-04ER54754

2. Divertor: Limiting Even for ITER 2007 ITER Physics basis: “The fusion gain in steady state maximizes at low density for constant N. The limitation on reducing the density in next generation tokamaks is set by the impact on the divertor.” The divertor sets the upper bound on the available power density in steady state mode Also: “It should be noted that presently developed advanced scenarios have not yet provided fully integrated scenarios and several issues remain to be solved, such as edge compatibility with the divertor” • For ITER (which has much lower power density than a reactor) • Advanced scenarios are already at or beyond the expected limit of standard divertor heat capacity

3. ITER-like divertor probably requires unacceptable core radiation in a reactor • Core fRad ~ 85% qualitatively different from ITER value • Required confinement of net power probably too high* • H89P ~ 4.2-5.2 • Burning plasma: Thermal instability* • Strongly self heated plasma plus high fRad probably leads to virulant thermal instability • Burning plasma: He build-up & Collapse* • High core He generation, low core transport likely leads to Helium collapse for fRad ~ 85% • CTF’s and hybrids are in between ITER and a reactor * Kotschenreuther,Valanju and Mahajan, Phys. Plasma July 2007

4. NEED a Divertor solution for much higher upstream q|| • With the same core radiation fraction as ITER, and ~ 5 times greater P/R, Reactor q|| is ~ five times greater • Steady state fusion power optimizes at lower density for current drive and lower edge density for bootstrap Need divertors which can handle higher q||and a low n • Also, ELMs/disruptions in a reactor would be • Several times larger than ITER • Hence, a divertor which can tolerate a much larger ELM/disruption is highly desirable

5. Problems at high divertor power/ low density • Divertor plate heat flux • Technological limits of ~ 10 MW/m2 , perhaps less at much higher neutron fluence than ITER • Helium pumping • In simulations, degrades very rapidly with power and lower density • Plasma erosion of the plate/plasma impurities • High plasma plate temperature/low density greatly increases sputtering/reduces prompt redeposition • Divertor survival of disruptions/ELMs • Most Solutions of 1 help 4 but don’t improve (or even degrade) 2 and 3 Need to scope out how bad the integrated problem probably is at DEMO/CTF parameters Need guidance on how to improve proposed solutions to get integrated solution for the devices we need to build

6. Research Plan • Start with relatively inexpensive simulations of innovative divertors for expected conditions of DEMO and CTF (both ST and normal Aspect ratio) • Examine integrated performance: Heat flux, helium exhaust, erosion, transients • See where the various innovations stand on all these variables, where improvement is needed • Examine: Standard divertor, highly tilted plate, Super X-divertor/ X-divertor, Snowflake divertor, liquid metal (Li, SnLi, etc.) and combinations • Improve edge simulation models as time goes on (better plasma transport models, dust/foam impurity generation, etc.) • Develop experimental tests of novel divertors on existing tokamaks (US and world) • Diagnostic upgrades to measure all relevant quantities • Compare experimental results with simulations • Testing must emphasize solutions which will probably extrapolate to a reactor- benchmarked simulations will be critical in determining this

7. SXD: saves NHTX from heat flux menace according to simulation • SOLPS 2-D calculations (Canik & Maingi) • With SXD & 30 MW, peak heat flux < 10 MW/m2 • Not possible with standard divertor (peak stays at 30-40 MW/m2) SOLPS SXD Calculation NHTX Standard Divertor NHTX SXD (Corsica Equilibrium)

8. Why the integrated divertor problem is harder (1): Erosion/Impurities • Temperature on plate (sputtering) increases rapidly with higher q|| , prompt re-deposion decreases (strongly) with lower density Plate tilting/flux expansion DO NOT reduce plate temperature or increase plate density • Temperature at plate is determined by q||, NOT BY Qplate • Most proposed divertor innovations reduce Qplate, BUT NOT q|| • SXD advantage: Only method of geometrically reducing q|| at plate (by going to larger R) • SXD: reduces Tplate by order of magnitude and increases nplate in simulations • Expect SXD to greatly reduce erosion and impurities

9. Why the integrated problem is harder (2): He exhaust ITER simulations: with increasing PSOL at fixed nupstream He exhaust gets worse even faster than heat flux Qplate ~ P2.5 n-2.4 --- but core He density ~ P3.3 n-5.5 • Most strategies to increase wetted area (plate tilt, flux expansion) probably degrade He exhaust • Recycled neutral He from plate must transport through the plasma without being ionized to reach the privated region and be pumped out • Spreading out the wetted surface (at constant Rmajor) reduces He reaching private region Higher PSOL together with higher plate tilt/flux expansion is likely to be a disaster for He exhaust • SXD advantages: • Very large reduction in Tplate/increase in nplate • For a given wetted area, reduces the distance He must go to get to the private region • Divertor is closer to outside world, so geometry might allow much better pumping

10. Simulations are an inexpensive way to look into the future to see which innovations will extrapolate to an integrated solution Shortcomings found in simulations can motivate improvements in the concepts where they are weak, and consequent experimental tests • Uncertain parameters (e.g. SOL width) can be varied within plausible range to indicate the robustness of various solutions • Simulation models will be improved as other thrusts bear fruit • Coupling to core simulations can show the degree of improvement in overall performance by enabling • Lower edge density (higher current drive, bootstrap current=> higher , E, etc.) • Lower core radiation with improved confinement, helium exhaust, thermal stability, etc.

11. Simulations for combinations of ideas • The divertor problem is so hard that combinations of innovations may be needed Example: combine novel geometries with fast flowing Liquid Metals (LM) or capillary/pore LM • Flux expansion can greatly reduce MHD drag effects for fast flowing LMs • SXD reduces drag several times further • SXD gives much better isolation of the core from evaporated impurities (because of the very long throat) Simulations can inexpensively examine the complicated physical processes in such combinations to find what works best • Experimental tests of extrapolatable combinations

12. Experiments • Obviously will benchmark, and be interpreted using, simulations In addition, SOL physics may be clarified by finding how the SOL changes as the magnetic geometry changes • Many innovative divertors vary important parameters far from a standard divertor (e.g. magnetic shear, line length, B and q|| at strike point, recycling (Li), etc.) Finding how the plasma responds could elucidate SOL physics, test theoretical models, etc. • Diagnostics need to be improved for the integrated problem: • Heat flux, plate temperature/density, plasma radiation, etc. • Possible He exhaust tests with outboard edge He fueling?

13. Potential Experimental Tests NEAR TERM • Reduced size plasmas on DIII-D with existing hardware • Snowflake being considered for this year, SXD next • Testing on MAST • SXD physics design finalized, engineering design in progress • Testing on SST • SXD is planned-possible opportunity for collaboration with diagnostics? • Tests of X-divertor (flux expansion part of SXD) on NSTX LONGER TERM, more expensive upgrades • DIII-D tests with baffling included for SXD • NSTX with coil/baffling/vacuum vessel modifications for SXD • NHTX/Vulcan • CTF (ST-CTF, FDF) and possible future hybrid • Pure fusion DEMO

14. Divertor challenge is more than plasma physics: SXD engineering advantages/disadvantages • Divertor is closer to the outside world, so maintenance might be much easier • Potentially much easier pumping for same reason • Divertor plates have substantial neutron shielding (several times less dpa and much lower energy => less He generation) Could substantially ease divertor material development

15. Divertor challenge is more than plasma physics: SXD engineering advantages/disadvantages • Disadvantage: vacuum chamber has a more complex shape • But calculations up to now show that the SXD DOES NOT REQUIRE a larger TF coil radius • Magnetic control requirements need to be refined • Preliminary analysis finds SXD is no harder to control than standard divertor • Magnetic geometry response to ELMs, disruptions, etc. The magnitude of engineering advantages/disadvantages for innovative divertors (including LMs) needs more thorough system assessment

16. Opportunity for US leadership • Focus of other major new experiments in the world is on areas other than Divertor and PMI innovations EVEN THOUGH THESE ARE THE AREAS WHICH ARE MOST SERIOUSLY IN NEED OF IMPROVEMENT TO MOVE FORWARD IN FUSION • By emphasizing investigations in these areas, US could establish leadership in a critical domain of fusion research The severity of the exhaust problem seems to be widely underestimated; simulations could help correct that deficit • Simulations can demonstrate the need for these innovative ideas in which the US already leads the world (SXD, snowflake, Lithium/Liquid Metal, etc.) Finally, undertaking concrete experiments to solve this very hard set of problems is bound to cement US leadership

17. Solution:Super X Divertor: evolution from X-Divertor XD/snoflake/tilt plate: expand flux SXD: maximally increase Rdiv SXD greatly increases: wetted area, line length, margin from sheath limited regime

18. SXD: Developed for both copper coil and superconducting devices Cu coil devices: • ST-CTF (Peng) A= 1.5 • Normal A CTF (FDF) A = 3.4 • NHTX A = 1.8 • Other ST-CTFs: A = 1.8, 2.5 Superconducting reactor designs • ARIES-AT A = 4 • Slim- CS A = 2.6

19. SXD Examples (CORSICA equilibria) With 5% less Amp-m than SD

20. SXD Examples (CORSICA equilibria)

21. SXD fits in TF corners - no TF real estate issues • For NHTX, FDF, and Reactors, SXD does not require larger TF coils • SXD uses available space (in the corner of TF coils) which is normally unused TF NHTX (PPPL/ORNL) CORSICA Equilibrium for NHTX-SX

22. SXD design space is smooth and large SXD is insensitive to plasma changes possibly because the long SXD “leg” near the SXD coils is far from plasma current

23. Gains for SXD equilibria

24. 1 degree plate tilt limit - limits wetted area • Minimum angle between total B and plate ~ 1 degree, (generally accepted limit) • Limits Aw gains for all flux expanders • Tilting plate, XD, Snowflake … • SOL width at midplane (so Asol) is given • So only knob left is increasing Rdiv • Pleasant surprise: SXD can be made with small changes in standard poloidal coil positions and currents (±5%) for a wide range of plasmas (low-A, high-A..) , HPDX - CORSICA Equilibrium

25. Avoiding the Sheath limited regime • Use analytic criterion for sheath limited regime (modified from Stangeby): Q||div / ne1.75L||0.75 = Q|| up (Bdiv / Bup)/ ne1.75L||0.75 • Since flux tube area ~ 1/B, Q||div ~ Q|| up (Bdiv / Bup) • Q|| up is determined by upstream SOL width, and factors independent of the divertor By increasing line length and decreasing (Bdiv / Bup) the SXD decreases the criterion by ~ 5 times

26. Neutron damage to divertor - critical issue • Tungsten “armor” on a high thermal conductivity actively cooled substrate • High conductivity substrates (Cu or C) severely deteriorate after only a few dpa • Reactor walls receive ~ 30-100 dpa • Only hypothetical high heat flux divertor materials might tolerate ~ 30-100 dpa • High risk, many decades away with much material development effort • Slow development would hamstring any high duty cycle DT device (CTF, DEMO) • SXD: substantial shielding of divertor plates for future CTF, DEMO Initial MCNP calculations- damage rate reduced by 10-20 times With SXD, much ITER divertor technology might by transferred to CTF, DEMO

27. Further SXD issues to be investigated • Doesn’t inner divertor now become the problem? • No. Almost all HPDs are designed as double-null (DN) devices • In DN, heat flux on inner plate is order of magnitude less than outer • SXD does not increase heat flux on inner plate • So unless SXD decreases heat flux by more than ~5, this is no problem • Won’t some instability increase cross-field transport? • That would be good! It will further increase SXD capability. • The SXD leg does not go very near the extra X-point, so ergodic problems (hot spots etc.) are expected to be not bad • What about pumping, He pumping, etc. • SXD is better isolated from plasma, lower T, higher n, so higher neutral pressure • TF ripple at SXD plates? Solution: shape SXD plates • Is stable detached operation possible with SXD? Need to simulate.

28. Broad interest in implementing SXD • SXD is now the “SD” for high-power density low-A devices (e.g., CTF) • SXD is now prominently featured in CTF presentations and actual engineering design work • PPPL-ORNL-IFS (SOLPS runs) working with SXD for NHTX • Preliminary results are very encouraging: SXD may be necessary and sufficient for NHTX • SXD is now integral part of PPPL plans for NHTX • Actual design to test SXD (at max Rdiv allowed by vessel/ports) is now underway in NSTX upgrade • MAST now has a working group to make SXD a part of MAST upgrade • A MAST-IFS collaboration has begun • IPR (India) has formed a group to design SXD on SST (1000 sec. SC tokamak) • Interestingly, SST heating power is currently limited by divertor - SXD will open new regimes for SST • PRC-IFS collaboration now exploring SXD for HL-2A • Preliminary designs being generated for the device • Very positive feedback from many at MIT meeting Feb. 08 • Common statement: SXD should be tested on some device as soon as possible

29. A possible HPDX with SXD Two possible missions: • Component Test Facility: short-term, useful, necessary fusion task (parallel to ITER) using conservative physics and technology, or • Inexpensively develop advanced physics modes for a pure fusion reactor in an integrated fusion environment with high heat fluxes

30. A 100 MWfusion HPDX (CTF) with SXD • Goal: get high power density but stay conservatively well within demonstrated physics and technology limits so it can be built soon (at least in principle) • For small size and coil mass, and easy maintenance: use Cu coil ST: A = 1.8 R = 1.35m = 3 7 T on coil Neutron wall load ~ 0.93 MW/m2 • enough room for neutron shield (~0.1 m): centerpost lifetime 1-2 yr

31. HPDX: dimensionless physics parameters • Two key parameters: <β>N = (<p>/<B2>)/(I/aB) and H [confinement enhancement over ITER98H(y,2)] • No-wall stability limit is βN ~3 for all A • Use estimated current drive efficiency I neR/Ph= 0.3 x 1020 (<Te>/10kev) A/Wm2 • very close to PPCS studies, ITER analysis • Use VMEC equilibria with fixed temperature and density profiles (characteristic of low-collisionality hybrid H-modes) to get βN , bootstrap current, and fusion power • Use Sheath limited regime parameter benchmarked to SOLPS S = Q||u (Bdiv/Bu) /n1.75 L0.75 / 2x10-27 > 1 => Te > 100 eV

32. 1020 HPDX: conservative design Take N = 3, Pfus =100 MW • Current drive power needed much lower at lower density • Only SXD has S<1 so plasma cools near plate • Need SXD to handle power exhaust at low density • Density ~ 1/3 Greewald

33. HPDX: a conservative design HPDX is within general expectations for experimentally accessible plasma shapes Low-A allows high k and we have such HPDX equilibria

34. A (still-improving) HPDX Corsica Equilibrium Iplasma = 11 MA, Btor = 2.5 T, Rmajor = 1.35 m, aminor = 0.75 m, Aspect A = 1.8 Elong  = 3.142 Beta = 19.1% Max Aw = 9 m2 Bcenter post = 7 T MCNP shows: neutron damage to Center post (with 10 cm steel cladding), and divertor plate can be kept below 2 and 1 dpa/FPY, so can operate for ~ 1-2 FPY

35. Proposed Next Step steady state devices-much more challenging divertor operation than ITER