Addressing ‘Knowledge Gaps’ and ‘Show-Stoppers’ in the ... · Thrust #1: Address the ‘knowledge gaps’ in edge transport physics: Develop fundamental understanding of the

Addressing ‘Knowledge Gaps’ and ‘Show-Stoppers’in the Edge Plasma Boundary

B. LaBombard

Towards building a credible vision for a DEMO-class fusion reactor:

Research Needs Workshop, UCLA, March 2-6, 2009

(1) The content of this talk is derived largely from the work of others;

I merely stand on the shoulders of giants.

(2) This is a ‘cheer-leading’ talk. Many comments/conclusions are obvious;

if not, they should be.

Inspiration from: R. Goldston, M. Kotschenreuther, T. Rognlein, P. Stangeby, D. Whyte,…


Fact: Power-producing fusion reactors will not be realized…

Example: The ‘power-handling gap’ -- the destruction of PFCs owingto plasma heat fluxes greatly exceeding material limits• Current experiments ~ not a problem; full operational space is

accessible, yielding excellent facilities for science exploration.

• ITER ~ severe ‘power-handling gap’; reliable disruption/ELMmitigation REQUIRED; detached divertor REQUIRED; operationalspace impacted; science mission potentially affected.

• DEMO ~ impossible(?) ‘power-handling gap’ -- factor of 4-5 times heatoutput as ITER (+ steady state + high temperature + neutrons…)

=> How do we build a credible fusion reactor concept?

…unless ROBUST solutions are demonstrated to CONTROL the magnitude of Plasma-Wall Interaction

this requires a shift in fusion research priorities


‘Power-handling gap’ ~ Lessons learned from ITER

Heat flux power widths at outer midplane from empirical scalings1:

p~ 5 mm (QDT=10 plasmas); p scaling with major radius.

• ~40MW/m2 on divertor without detachment (Prad ~ 20%)=> melting of divertor plate (W) with ~1 s exposure.=> impurity seeding, detachment required (Prad ~ 70 %).

• ~ 10 MJ/m2 Type I ELM energy deposition => 20x damage threshold! ELM mitigation required.

[Note: RMP ELM-suppression was unknown when ITER was designed!]

1Loarte, A., "Power and particle fluxes at the plasma edge of ITER : Specifications and Physics Basis," 22nd IAEA.2Loarte, A., et al., J. Nucl. Mater. 266-269 (1999) 587.

But… p in the divertor is uncertain; empirical scalings of divertor heat fluxfootprints are ambiguous2 -- covariances in the data set; no scaling withR? Two root problems: (1) We have no fundamental understanding ofthe physics that sets the power width; (2) Our magnetic confinementscheme focuses plasma exhaust onto a small target.


CFC4: 100xELM, 1.5 MJ/m2

WL2: 100xELM, 1.0 MJ/m2

W4: 100xELM, 1.5 MJ/m2

WL3: 5xDisruption, > 2.2MJ/m2

CFC5: 5xDisruption, > 2.2

MJ/m2

ELM simulation results fromQSPA facility in Troitsk1

Solid divertor target materials are already operating attheir limits for heat-flux removal

1Klimov, N., et al., "Experimental study of PFCs erosion under ITER-like transient loads at plasma gun facility QSPA," 18th PSI Conference, Toledo, Spain, 2008.

ELM duration ~ 0.5 msELM size ~ 1/10th ITER!

Exposed samples (preheated to 5000C)


Pure tungsten and lanthanum tungsten (comparison)

0.8 1.00.9

1.3 1.4 1.6

W(>99,96%),Qmax=1MJ/m2

(100 pulses)

Negligible mass

loss

W-1%La2O3,Qmax=1MJ/m2

(100 pulses)h= 0.04

μm/pulse

W (>99,96%),Qmax=1.5MJ/m2

(100 pulses)h= 0.06

μm/pulse

Plasmastream

direction

WL2

Absorbedenergydensity,MJ/m2


1Klimov, N., et al., 18th PSI Conference, Toledo, Spain, 2008.

Results from QSPA facility1


PFCs erosion, μm/pulse

3-0.040.005W-1%La2O3

-0.06negligiblenegligibleW (>99,96%)

30.3negligiblenegligibleCFC

>2.21.51.00.5Energy density,

MJ/m2

DisruptionELM Type I


Results from QSPA facility1 CFCs aren’t the solution either

ITER has legislated a 0.5 MJ/m2

limit for all ELMs

1Klimov, N., et al., "Experimental study of PFCs erosion under ITER-like transient loads at plasma gun facility QSPA," 18th PSI Conference, Toledo, Spain, 2008.

ELM/disruption suppression is required for ITER

ELM/disruption suppression is required for a DEMO


Steady-state heat fluxes (no ELMs) in C-Mod already exceed material limits


Axisymmetric melt damage to Moly divertor tiles from ~0.2 s of 1 discharge:~4 MW into SOL (enhance L-Mode) -- P/S ~0.6 MW/m2

q// ~0.7 GW/m2 + operator mistake, strike-point on pumping slot tiles!


Robust† solutions (if they exist) will be found only by pursuing twoessential, complementary research thrusts:

Thrust #1: Address the ‘knowledge gaps’ in edge transport physics• Among other things, we need to understand the physics that sets the SOL

power widths, with reliable scaling to ITER and DEMO…How bad is the‘power-handling gap’ really going to be?

• Can we exploit some characteristics of edge transport to spread heat fluxes?What happens if we employ different boundary-layer magnetic geometries?

Advanced Materials are not going to bridge the ‘power-handling gap’ for a DEMO-class device

†Robust = accommodates full plasma exhaust over a wide operational space of the device with a large margin of safety







Thrust #2: Explore & develop advanced boundary-layer/divertor systemsthat can truly ‘tame the plasma-material interface’ in a DEMO

• Increased core/edge radiation can help…but it’s not a robust solution1

• Employ brute-force approaches: magnetically expanded divertor ‘footprint’,enhanced divertor radiation, advanced target concepts including liquids…


1Kotschenreuther, M., Valanju, P.M., Mahajan, S.M., and Wiley, J.C., Phys. Plasmas 14 (2007) 072502.







Success in these two areas would address many urgent PFC issuessuch as component lifetime, impurity control, dust control...

1Kotschenreuther, M., Valanju, P.M., Mahajan, S.M., and Wiley, J.C., Phys. Plasmas 14 (2007) 072502.


Thrust #2: Explore & develop advanced boundary-layer/divertor systemsthat can truly ‘tame the plasma-material interface’ in a DEMO

• Increased core/edge radiation can help…but it’s not a robust solution1

• Employ brute-force approaches: magnetically expanded divertor ‘footprint’,enhanced divertor radiation, advanced target concepts including liquids…


Thrust #1: Address the ‘knowledge gaps’ in edge transport physics:Develop fundamental understanding of the physical processes

that control edge/divertor plasma transport

(1) Fully characterize boundary plasma phenomenology and its response to a broad range of‘external’ controls, particularly targeted for a DEMO (power, LCFS topology, divertorgeometry, fueling, RMP,…)Edge profiles: ne Te, Ti, , nz (Edge = pedestal and SOL)

Edge turbulence, critical-gradient transport dynamics, (blobs, ELMs, micro-turbulence)

Edge flows (SS & intermittent), flow shear profiles, interaction with micro-turbulence

Mapping Pedestal-to-Divertor: heat, particles, turbulence,…

Divertor response: detachment, neutrals, opacity, radiation, impurity entrainment,…

Plasma-sheath physics (non-thermals, impurities, neutrals,…), including material response

Goals/Approach:




(2) Build and test physics-based empirical models which embody plasma responseTight coupling between theory and experiment: identify and map dimensionless phase-space

Wind-tunnel similarity experiments; ability to accurately ‘interpolate’ among experiments, withminimal extrapolation distance to DEMO







Goals/Approach:




A quasi-DEMO device (hybrid of Vulcan, NHTX, FDF…) is required to assemble this knowledge base.









Goals/Approach:




(3) Build, refine and (de)validate test first-principles numerical modelsTight coupling between experiment and theory; iterative refinement of numerical simulation

Guidance from dedicated, small-scale plasma turbulence experiments

Time-averaged transport codes coupled to first-principles turbulence models

=> Ability to predict plasma response and optimize boundary plasma ‘solutions’ for a DEMO



A quasi-DEMO device (hybrid of Vulcan, NHTX, FDF…) is required to assemble this knowledge base.







Goals/Approach:


Experimental Facilities ~ Near Term• We have excellent confinement facilities to assess key knowledge gaps

C-Mod: High P/S, high neutral opacity, W/Mo,…

DIII-D: Flexible topologies, diagnostic access, C,…

NSTX: ST topologies, Li,…

$$ Yet, the boundary research programs are lacking the personnel, diagnostics and programmatic support to carry out this mission.

Tools Required:









Tools Required:



Experimental Facilities ~ Longer Term $$$ We need to aggressively pursue boundary plasma science in a new confinement facility that can truly simulate the DEMO environment.

• DEMO-level P/S (~1MW/m2); non-DT with DEMO topology is preferred

• Large, flexible divertor geometry to develop/test innovative divertor solutions


Theory & Modeling• A tight coupling among experiment-theory-modeling is essential! Yet progress

in experiment-simulation comparisons has been painstakingly slow.

$ We need increased resources (personnel, programmatic focus) dedicated to this, including on-site theory/modeling support, working with experiments.






Tools Required:



Experimental Facilities ~ Longer Term $$$ We need to aggressively pursue boundary plasma science in a new confinement facility that can truly simulate the DEMO environment.

• DEMO-level P/S (~1MW/m2); non-DT with DEMO topology is preferred

• Large, flexible divertor geometry to develop/test innovative divertor solutions




Output:

(1) Quantum-leap in boundary-layer plasma science

(2) Reliable extrapolations to the edge plasma state in a DEMO:

• Power channel widths and heat flux footprints

• Compatibility of ‘power-handling solutions’ with core (pedestal)

• Particle/energy fluxes to wall surfaces

• Impurity behavior (redeposition, screening, transport)

• Helium exhaust, plasma fueling and pumping




Yet, improved physics understanding alone will not solve the critical PWI

issues for DEMO, including the ‘power-handling gap’…

Output:

(1) Quantum-leap in boundary-layer plasma science

(2) Reliable extrapolations to the edge plasma state in a DEMO:

• Power channel widths and heat flux footprints

• Compatibility of ‘power-handling solutions’ with core (pedestal)

• Particle/energy fluxes to wall surfaces

• Impurity behavior (redeposition, screening, transport)

• Helium exhaust, plasma fueling and pumping


1Najmabadi, F., et al., Fusion Engineering and Design 80 (2006) 3.

ARIES-AT1

ITER

Psol~80 MW Psol~250 MW

Keeping all else the same,core/pedestal radiation fraction mustscale up dramatically.

Yet, core confinement is tied to thepedestal conditions (temperature).

The ‘radiation solution’ trades-off coreperformance for divertor powerhandling. Current experiments havenot been forced to make this trade-off…

Plasma exhaust is magnetically focused toward asmall ‘footprint’ on the divertor target

… How well would today’s tokamaks perform if they were constrained toalways radiate 80% of the input power from the confined plasma region?

Fundamental Problem:The volume apportioned for plasma exhaust in today’s tokamakexperiments is much too small for a DEMO-class device


1Mahdavi, M.A., et al., Fusion Science and Technology 48 (2005) 1072.

Doublet-III ~1981Expanded Boundary1

DIII-D ~1984 DIII-D ~1994

Legacy of optimizing a device’s cross-section to maximize theconfined plasma performance --- Very appropriate, until now...

Evolution of Doublet/DIII-D cross section

Fundamental Problem:The volume apportioned for plasma exhaust in today’s tokamakexperiments is much too small for a DEMO-class device


Thrust #2: Explore & develop advanced boundary-layer/divertorsystems that truly ‘tame the plasma-material interface’ in a DEMO

DEMO requires us to ‘rethink the box’ and explore innovative divertors,expanding the plasma boundary as needed to obtain robust solutions.

1Pitts, R.A., Chavan, R., and Moret, J.M., Nucl. Fusion (1999) 1433.2Ryutov, D.D., Cohen, R.H., Rognlien, T.D., and Umansky, M.V., Phys. Plasmas 15 (2008) 092501.3Kotschenreuther, K., et al., 22nd IAEA, 2008.

TCV cross-section1

• Increase plasma-wetted area of the divertor target plate

• Explore advanced materials, including self-renewing surface concepts such as liquids

• Increase radiated power in the open field-line region

Expanded Boundary(EB)

Snowflake2

(SF)Super X Divertor3

(SXD)



Magnetic-expansion concepts can offer the robust solutions that we seek.

• Reduction of target flux densities in all operational regimes

• Reduced target neutron fluence

• Excellent pumping and impurity entrainment

• Shielding core from divertor target events (UFOs)

• EB and SF appear ideally-suited for convention tokamak

• Extreme magnetic shear in SF -- spreads heat fluxes?

• transport in long divertor leg geometries (SXD)?

• transport when B-field is ~tangent to surfaces (EB, SF)?

• Use of self aligning surfaces (liquids, ‘fire-polishing’) withEB and SF concepts?

• Use of stochastic magnetic field structures?

Exciting issues to explore…



Magnetic-expansion concepts can offer the robust solutions that we seek.

$ Continued concept development and refinement

$$ Vet promising ideas in current confinement devices

Challenge:

Use plasma-transport physics and wall-interaction dynamicsto dissipate ~250 MW of plasma efflux, steady state, at hightemperature, with high reliability and long component lifetimein a DT fusion nuclear environment.

High Priority Tasks:

$$$ The first priority of any new quasi-DEMO facility (Vulcan, NHTX, FDF,..) must be to aggressively test, diagnose, and demonstrate a robust boundary- layer/divertor system that will truly ‘tame the plasma-

material interface’your idea here

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Addressing ‘Knowledge Gaps’ and ‘Show-Stoppers’ in the ... · Thrust #1: Address the ‘knowledge gaps’ in edge transport physics: Develop fundamental understanding of the