Latest Results: Task IV, FED paper, etc.

1. Latest Results:Task IV, FED paper, etc. Investigators • IFS: M. Kotschenreuther, P. Valanju, L. Zheng • LNLL: T. Rognlien

2. Four Areas Considered • Effect of fast flowing LM wall on plasma MHD modes • Summary of submitted FED paper: Reactor Implications of convective SOL transport • In collaboration with Tom Rognlien; use2-D simulations (UEDGE) & IFS neutral code NUT • ICC funding grant awarded for APEX developed concept: “field line extraction divertor” • Develop coil designs for reactors and retrofit experimental tests • The importance of an APEX-like program in the future

3. Effect of a Fast Flowing Wall of Plasma MHD Modes • Plasma MHD code has been completed • Has been benchmarked against GATO for circular cross section plasmas • Code works for non-circular plasmas, but coupling to the vacuum/resistive wall has only been completed for • circular cross section • large aspect ratio plasmas • Effect of a fast flowing wall on pressure driven kink modes in a high beta plasma (circular) presented here • Even though circular, these results are a large improvement in realism over previous analysis

4. Recall Previous Results • Previous analysis by Zakharov, Kotschenreuther • Low betaN, not high betaN like a reactor • Different instability drive mechanism from reactor cases: plasma current rather than plasma pressure • Different plasma current profile from reactor cases • Most analysis also circular • Only qualitatively similar to reactor cases • New Results: • Realistic high bootstrap fractions considered • Reactor relevant high betaN from 3 to 6 considered • ARIES RS has betaN = 4.8 • Significant plasma profile optimization • But still only circular

5. Results for Realistic Plasma Profiles BetaN = 3 case • Stabilization for betaN= 3 : • For 2 cm Li, flow > 20 m/s • For 2 cm Sn, flow ~ 50 m/s • Flow velocity increases sharply with betaN • For betaN = 4-5 (ARIES ) flow > 100 m/s (even for separation = 0) • PRELIMINARY RESULT: realistic LM flow velocity will not stabilize desirable cases with high beta and high bootstrap fraction • BUT: Elongated Plasmas might give better results

6. Two Routes to High Beta • Higher Elongation • High temperature edge boundary condition (even for low elongation): using Li edge or extraction divertor • We have examined case 2) here for circular plasmas • For beta = 10 % (ARIES AT) circular plasmas have higher betaN => HARDER TO STABILIZE RWM • Elongation=2.2: betaN = 5.6 • Circular betaN = 8 • For betaN > 3.7 resistive wall modes are very severe • cannot be stabilized even for Li flows > 100 m/s (2 cm thickness) • cannot be stabilized with feedback (n numbers too high)

7. Future Investigations • Non-circular geometry (high elongation) • Controlled alpha loss to cause MHD stabilization by rotation (Have modified MHD code to include rotation; almost complete) • Controlled alpha loss solves multiple problems: • Plasma MHD stability from flow: verified in experiments • Control of thermal runaway in plasma AT modes with internal transport barriers • Alphas hit wall with Mev energy so liquid pumping is much more efficient • Alpha loss concept requires liquid surface for erosion control, but is more flexible: • Works with flibe, slow moving LMs • May only require very limited wall coverage with a liquid • May only require a thin liquid wetted surface for erosion/pumping, rather than a fast stream

8. EFFECTS of CONVECTIVE BLOB TRANSPORT on REACTORS • Subject of submitted FED paper: summary here • In collaboration with Tom Rognlien, Prashant Valanju

9. Our Physics Understanding of SOL Transport has Changed • Pioneering experiments at C-Mod and DIII-D: large convective transport of plasma blobs • Theoretical investigations: blobs of plasma should rapidly convect to the main chamber wall • IFS-LLNL collaboration: investigation of potential REACTOR effects of convection for the first time

10. Serious Effects of Blob Convective Transport • First Wall Erosion • A concern mentioned in the literature, but heretofore not estimated for reactors • Serious implications found • Helium Pumping • Not discussed in literature, but: far SOL transport disproportionately effects helium removal--work in progress

11. DIII-D Simulation (Pigarov) 100 Convective Velocity m/s IFS Reactor Simulation DIII-D Simulation (Pigarov) Distance to wall 2-D Simulations using UEDGE • Present state of the art: use empirical diffusion coefficients for reactor simulation • But convection appears essential in far SOL • We use empirically motivated convection model similar to that on present experiments to estimate reactor effects

12. Several Convection Profiles Examined • Convection profile varied so that the simulated SOL density profile for a reactor matches SOL profile characteristics found in experiments • Quantitative profile characteristics checked with experiments: • Ratio of density in the SOL (d/a=0.04) to separatrix density • Density scale length at d/a = 0.04 • Convective flux at d/a = 0.04 • Four convection profiles tried: A) 10 m/s to 100 m/s (found to give BEST FIT to data) B) 10 m/s to 50 m/s (reduce convection near wall) C) 5 m/s to 100 m/s (reduce convection near plasma) D) 0 m/s to 0 m/s (no convection) • Case A gives best match: reducing the convection (B and C) results in a poorer match to characteristics 1,2 and 3

13. With conv No conv Why Wall Erosion Estimates Based on Models w/o Convection are Likely to be Low • Standard SOL transport model: constant diffusion only • Probably underestimates plasma-chamber interaction by ~ 30

14. Density Ratio Comparison • Find: a strong relationship between the density ratio in SOL and the Greenwald ratio • For high Greenwald ratio (like reactors), density in the far SOL is high => STRONG WALL INTERACTION • Case A is most consistent with experiments • Case D without convection does not match experiments Density Ratio = nsep/nd/a=0.04

15. Wall Flux Comparison • Experimentally estimated plasma flux to the wall has a large scatter • Flux trends are described by the expression of Labombard: GL=1021(ne/1020)2 • Case A most similar to data • Case D without convection: under-estimates flux by nearly two orders of magnitude

16. SOL Density Scale Length Comparison • SOL density decay is slow in experiments, and tends to be flatter at higher density • Case A most consistent • Case D without convection: density profile does not match experiments

17. Kinetic Neutral Code NUT evaluates the hot CX neutral flux to the wall • Wall erosion is dominated by hot CX neutral flux to the wall • The fluid treatment in UEDGE cannot evaluate this • Thus: the kinetic neutral code NUT is used • NUT: benchmarked against experimental data on TEXT & C-MOD • The plasma profiles and neutral source found by UEDGE are input into NUT • NUT computed the energy distribution of CX neutrals back the the wall • The sputtering coefficient is integrated over the CX neutral distribution to obtain the wall sputtering

18. With Realistic Convection: Strong First Wall Erosion • Most sputtering-resistant material: Tungsten • Without convection: 0.17 mm/yr • With convection: 0.61 mm/yr • estimate small prompt re-deposition

19. Consequences of Tungsten Erosion • Large dust generation • ITER: ~10% of sputtered material forms micron dust • With convection: ~ 340 kg/yr dust after 2 years • LOVA dosage marginally exceeds no evacuation limit (even with 99% filter, -adapting analysis of Merril et. al.) • Plasma Impurities • C-Mod has high-Z wall: H-mode screening factors 1-10% • ASDEX ~ 1% UEDGE ~ 10% • This range of screening can have unacceptable consequences: • H-mode ignition precluded due to radiation for ~ 0.5 1.0 % penetration

20. Implications for Liquids • Flibe, LiSn, Sn considered • Obviously dust, structural erosion are not issues • For low Z PFCs (Flibe, LiSn): • Plasma: much more tolerant of Low Z impurity • Acceptable screening factor ~ 5 % • Recall experimental values are ~ 1 - 10 % • Sn walls • High Z: acceptable concentration slightly higher than W, but sputtering also slightly higher • Required screening factor ~ same as W: very worrisome

21. Implications • Better physics understanding of SOL transport required: could be show-stopper • Plasma-wall interaction: structural erosion, dust • Impurity transport and core plasma contamination • Alternative design concepts required • Low-Z liquid walls • Low Z => acceptable plasma impurity level ~ 1% • Continually replenish wall => no structural erosion • Extraction divertor • Low density SOL operation to minimize SOL convection

22. Beyond FED paper: Present Work • Erosion near edges of protrusions and cavities • Near corners, projections: blobs will dump plasma much more strongly • Recycled neutral source many times higher => Local erosion rates several times higher (?) • Wall next to ICRF antennas, and antenna itself • Wall near blanket test modules which are inset by ~cms • Assuming ITER edge is the same as previous calculation • 10,000 shots, 400 sec => ~ 3 mm Be erosion for flat wall • Several times higher (?) near protrusions and cavities • Reactor relevant design solutions (?) • Low-Z liquid wetted wall near edges? • Extraction divertor to run in SOL regime with low blob transport?

23. Future Work • 3-D NUT calculations to examine hot CX caused erosion near edges (with model SOL profiles) • Better models of SOL turbulence & simulations • More physics based models of blob erosion • Low density, high T edges have lower blob transport? • Investigate convection effects on He exhaust using UEDGE • He path from the core to exhaust is mainly in the far SOL • How seriously is He exhaust degraded by strong SOL convection?

24. Have Received ICC Grant to Design Extraction Divertor • Use design/optimization tools developed for NCSX Compact Stellarator • Highly sophisticated algorithms optimize coils for 3-D magnetic fields • Optimizations can include an arbitrary number of engineering and physics properties • After modification, tools will enable optimal coil designs to be developed for extraction divertors in many scenarios: • Reactor scenarios • Retro-fits of existing devices to test the concept (NSTX, Pegasus, CDX-U, others?) • Why these tools? • Battle-tested: greatly simplified NCSX coils • Range from initial “filament” to full engineering design • NCSX tools developed after much effort- allowing extended divertor design to be done with relatively small incremental cost to DOE

25. External Divertor Coil Design/Optimization • Similarity with NCSX Compact Stellarator coil design: • 3-D magnetic field requires optimized coils • Manual optimization has proved principle, but gives complex design • Design parameter space multi-dimensional • Both engineering and physics target functionals are required • Target functionals can be quantified • Targets are non-linear functions of input parameters • Possibility of many isolated optima • Sophisticated optimization methods are needed

26. Optimization Targets • Engineering targets (new and already in NCSX tools): • Clearance of extracted field lines from coils • Currents, (heating for Cu coils) • Stresses • Geometry: Curvature, Torsion, manufacturability • Effects of finite cross-section (filament -> real coils) • Feeders: geometry, stresses, and ripples • Ease of assembly and replacement • Neutron fluxes on coils • Physics targets: • Ripple in plasma • Plasma recycling • Plasma shaping: elongation, triangularity

27. Why Liquid Surface Investigations Should be Continued • In the past, APEX justified liquid surface work for long term objectives • A cheaper fusion reactor • But liquid surfaces may be needed even for lowest order feasibility • Erosion of divertor plates due to ELMS • Radiation collapse of the core plasma for W walls (reactor) • Structural erosion especially near edges • Dust generation and environmental/regulatory acceptability • The present strategy is to find a plasma operating regime which is compatible with all the constraints of solid walls, as well as • High plasma radiation fractions • Low disruptivity (antithetical with high radiation fractions) • Good confinement and high beta (antithetical with low temperature edge) • We should admit that a viable plasma operating regime may not exist, and that the engineering constraints may need to be alleviated through novel approaches

28. The Value of an APEX-like Program in the Present Environment • Perhaps instead of positioning APEX to develop futuristic reactors, it should be positioned as providing novel engineering solutions to zero’th order feasibility issues • The value in APEX is in providing options to avoid technological/physics dead ends which may be too difficult to solve • This value could be substantial even for next generation burning plasma devices • To make this value more apparent, it might be useful to write an analysis • Of conventional divertor/first wall options and their engineering risks • How those risks could lead to much wasted time/funds • How broader technological options examined by APEX could provide crucial alternatives

29. Why APEX-like Investigations Should be Continued – ITER Relevance? • Even for ITER, thin wetted liquid surfaces could greatly improve the chances for success by: • Eliminating erosion concerns from ELMS in the divertor • Eliminating erosion concerns on the main chamber first wall without introducing the plasma contamination concerns of W • A wetted divertor application requires a thermo-hydraulic & MHD analysis, the expertise for which resides in the chamber technology area • A wetted section of the first: also chamber analysis • It appears to me that a compelling case can be made for liquid surface work within the chamber technology area, even without the goal of a high neutron wall load • Liquid surfaces (perhaps thin wetted surfaces) for erosion/impurity control may be needed even for basic feasibility