Plasma Wall Interactions (PWI) Panel Introduction

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Plasma Wall Interactions (PWI) Panel Introduction

• J.N. Brooks1 and the
• ReNeW Theme III PWI-Panel
• 1Purdue University
• ReNeW Meeting, UCLA, March 4-6, 2009

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Plasma Wall Interactions (PWI) Panel

• ReNeW Theme III Taming the Plasma Material
  Interface
• Mike Ulrickson, Chair
• Rajesh Maingi, Vice-Chair
• Rostom Dagazian, DOE/OFES
• Plasma Wall Interactions (PWI) Panel
• Jeff Brooks (Purdue), Chair
• Jean Paul Allain (Purdue)
• Rob Goldston (PPPL)
• Don Hillis (ORNL)
• Mike Kotschenreuther (U. Texas)
• Brian LaBombard (MIT)
• Tom Rognlien (LLNL)
• Peter Stangeby (U. Toronto)
• Xianzhu Tang (LANL)
• Clement Wong (GA)

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ReNeW Meeting Inputs PWI

• PWI Conference Calls
• White Papers- 40 Theme III, 15 PWI
• Inputs from community

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Plasma Wall Interaction Panel

• PWI panel topic defined to cover
• Plasma edge, scrape-off layer
• plasma parameters, heat, particle flows
• First 1?m of plasma facing component surfaces
• 1-10 nm, for sputtering
• 1?m for micro-structure evolution, dust,
  bubbles, etc.
• 1?m for plasma transient response (e.g. vapor
  formation)
• Does not cover (but interfaces with)
• Plasma core
• Bulk material properties/effects (e.g. neutron
  damage, tritium permeation)

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Theme III White Papers
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Plasma/Material Interactions

• PWI Panel believes
• Plasma/material interactions is probably the
  single most critical technology issue for fusion.
  
• Concerns
• (1) Plasma facing component lifetime
• (2) Core plasma impurity contamination
• (3) Tritium inventory/operational requirements
• Critical Issues
• Sputtering erosion and impurity transport
• Plasma transient erosion (Edge Localized Modes
  (ELMs), disruptions, runaway electrons.)
• Plasma contamination (core/edge) due to erosion
• Tritium co-deposition in eroded/redeposited
  material, and mitigation
• Important Issues
• Dust-formation and transport safety
• For tungsten-He, D-T, bubble formation and
  effects
• Hydrogen isotope and helium trapping, reflection,
  etc.
• Mixed-material formation/integrity

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Fusion plasma facing material requirements

• Heat flux
• 10 MW/m2 peak (ITER, on divertor), normal
  operation
• 0.01 - 100 GW/m2 peak, w/plasma transients
• 100 MW (ITER) - 600 MW (commercial reactor)
  total surface heat load
• Particle flux
• D-T 1023 - 1024 m-2s-1 _at_ 1-1000 eV
• He2 1022 - 1023 m-2s-1 _at_ 10-1000 eV
• Ok 0.1 of D-T
• Neutron flux
• 0.5 MW/m2 (ITER)
• Other
• Pump helium at fusion generation rate (optional)
• Pump D-T (optional)
• Low to moderate neutron activation
• Note Surface coating material does not need
  excellent structural properties.

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Some examples of PWI Issues

• It is not clear if PFCs in ITER can survive even
  one major disruption
• Giant ELMs in ITER are not tolerable to C or W
  surfaces
• VDEs, runaway electrons pose very serious
  threats to PFCs
• W fuzz effects in ITER surface
  integrity/erosion
• Major issue for predicting convective edge flow,
  turbulence generally
• T/Be codeposition, cleanup
• For Demo-most of above issues highly uncertain
  heat/particle flux values, ability to handle
• Present machines Mo sputtering D retention in
  CMOD, NSTX Li boundary effects

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Be-W interaction can lead to extreme failure
(PISCES crucibles)
Intact W wall (97W, 3O)
Inner wall coating (4 W, 95 Be, 1O)
Be22W?
Crucible failure zone (9 W, 70 Be, 14 C, 7
O)Be12W?
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Candidate tokamak plasma facing materials

• High power/large-area components (Divertor, Wall,
  Limiter)
• Elements-Solid
• Beryllium
• Carbon
• Tungsten
• Misc. (B, V, Fe, Mo)
• Elements-Liquid
• Lithium
• Gallium
• Tin
• Misc. applications
• Diagnostic mirrors-Be, Mo, Au, Rh, etc.
• Antenna insulators, e.g. YO
• Low-activation compounds- SiC

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PWI Panel Typical Sentiment (R. Goldston)

• As I talk to folks around the community, I am
  frequently shocked by how poorly they appreciate
  how serious the PWI issue is. The lack of
  understanding combined with the lack of
  demonstrated solutions is extremely serious.
• If we don't have 80 bootstrap current, we can
  still make fusion energy. If we need 1.5 m thick
  90 enriched 6Li blankets because we got some
  cross-sections wrong, we can still make fusion
  energy. I don't think we have a solution to the
  PWI/PFC problem similar to these.

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PWI Panel Typical Sentiment (P. Stangeby)

• PWI places at risk the successful development of
  MFE in a number of potentially show-stopping
  ways, including destruction of the walls,
  unacceptably high contamination of the confined
  plasma and unacceptably high tritium retention.
  PWI is largely controlled by the plasma outboard
  of the separatrix .
• It is not surprising that understanding of the
  SOL is so incomplete there have been several
  orders of magnitude more effort invested in
  confinement physics than in SOL physics, although
  the SOL is a considerably more complicated
  problem than the main plasma.

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PWI Panel Typical Sentiment (B. LaBombard)

• in the area of boundary layer physics and
  plasma wall interactions these (knowledge) gaps
  are extreme.
• At present, we have no physics-based model that
  can accurately simulate the heat-flux power
  widths observed in tokamaks, let alone scale them
  to ITER and DEMO.
• we must explore innovative concepts that can
  truly tame the plasma-material interface
  systems that control cross-field heat/particle
  fluxes, expand the plasmas interaction area
  (footprint) with material surfaces, and lead to
  robust, plasma-wall interfaces with advanced
  materials, including liquid surfaces. Success
  would provide credible solutions to DEMOs
  power-handling gap and also address other
  urgent issues such as PFC lifetime, impurity
  control, dust production and control.

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R. Goldston and the
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Gaps As summarized in e.g. 1, these are
extensive gaps in existing PMI theory,
modeling/code efforts and experimental
validation, including 1. Analyzing/explaining
many existing results, e.g. CMOD Mo divertor tile
erosion results, enhanced plasma performance in
NSTX lithium shots, as well as for numerous
international machines (JET etc.) where the US
could make a substantial contribution. 2.
Modeling/analysis of scaling and intermittent
character of SOL turbulent transport determining
heat-flux and particle-flux profiles on PFCs
(divertor, walls), and subsequent impurity
transport back to core. 3. Mixed materials
(e.g. Be/W, C/W) plasma induced formation and
response. 4. Sheath wall near-tangential sheath
parameters (this being critically important in
ion acceleration and heat transmission), ICRF
induced sheath and effects for ITER and future
devices. 1. R. Goldston and the ReNeW PMI Panel,
PWI Gaps vs. Tools to Develop Understanding and
Control
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Gaps-continued 5. Liquid metal surface (Li,
Sn, Ga) response including He and D-T
pumping/reflection and effect of same on
edge/core plasma, temperature-dependent sputter
yields, sputtered/evaporated material in-plasma
transport. 6. Tungsten nanostructure changes
due to He, N, etc. 7. Dust formation and
transport. 8. Plasma transient effects and
resulting core-plasma operating limitations in
ITER and DEMO, and solutions to same. 9. Atomic
and molecular data-gaps in database. 10.
Hydrogen isotope retention in He and D-T
irradiated materials. 11. Supercomputing-There
is a general major need to develop/improve
stand-alone PMI supercomputer capability (in
particular via implementing OMEGA real-time
coupling) as well as to incorporate PMI code
packages into integrated (SCIDAC, FSP etc.)
projects.
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Erosion/redeposition analysis summary-ITER e.g.
1

• Some confidence of acceptable Plasma Facing
  Component performance
• Beryllium wall-sputter erosion rate appears
  acceptable (0.3 nm/s) (for low duty-factor
  ITER).
• Be wall-core plasma contamination appears
  acceptable (2 Be/D-T)
• Tungsten (outer) divertor (baffle/target) net
  erosion rate appears negligible.
• W core plasma contamination (from W wall or
  divertor) appears negligible
• Tritium codeposition in redeposited beryllium is
  a concern, but probably acceptable ( 2 gT/400 s
  shot)
• Be/W interaction at outer divertor may be
  acceptable (no net Be growth over most/all of
  divertor target).
• Micro-structure (fuzz) formation of
  wall-tungsten may be acceptable (for low
  duty-factor ITER).
• Major Uncertainties
• Plasma SOL/Edge convective (blob) transport,
  and plasma solutions generally.
• Sputtered impurity transport w/ convective
  transport.
• Mixed (Be/W, etc.) material properties.
• 1 J.N. Brooks, J.P. Allain, R.P. Doerner, A.
  Hassanein, R. Nygren, T.D. Rognlien, D.G. Whyte,
  Plasma-surface interaction issues of an all
  metal ITER, Nuclear Fusion 49(2009)035007.

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ITER outer first wall sputtering rates OMEGA/WBC
analysis, convective edge plasma regime

• Be sputter erosion acceptable for low
  duty-factor ITER will not extrapolate post-ITER
• W erosion very low
• Bare-wall erosion low

Key additional required work convective
transport model upgrades/use detailed spatial
resolution, inner wall analysis, wall sheath
effects, rf sheath effects
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Plasma Transient PMI analysis summary-ITER e.g.
1-2

• Some encouraging results
• An acceptable (no-melt) plasma ELM parameter
  window exists for a tungsten divertor.
• A dual-material option may ameliorate runaway
  electron damage.
• Major Problems/Uncertainties
• An unacceptable (melt) ELM parameter window
  exists for tungsten.
• A big part of parameter space for plasma
  transients would severely impact the PFC
  surfaces.
• Giant ELMs
• Other ELMs
• Vertical Displacement Events (VDEs
• Disruptions
• Runaway electrons
• 1 J.N. Brooks et al., Nuclear Fusion
  49(2009)035007 2 A. Hassanein et al., PSI-18
  (2008), J. Nuc. Mat. to be pub.

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HEIGHTS parameter window for W divertor
acceptable (no-melt) ELM response

• A safe-operation window exists for tungsten.
• Note Carbon does not melt, but ELM material
  losses not fundamentally different than tungsten.

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Erosion/redeposition analysis for DEMO (via rough
extrapolation from ITER analysis)

• -- Low-Z materials are unacceptable due to
  sputter erosion.
• -- Candidate materials high-Z, i.e., W (Mo?,
  etc.) wall divertor, liquid metal divertor (Li,
  Sn, Ga)
• Some encouragement
• Tungsten divertor (baffle/target) net erosion
  rate and core plasma contamination rate from
  divertor could be acceptable.
• Tungsten wall sputtering erosion and core plasma
  contamination could be acceptable.
• Tritium/tungsten codeposition likely to be
  acceptable.
• Major Uncertainties
• Plasma SOL/Edge convective (blob) transport,
  and turbulent plasma solutions generally
  heat/particle-loads.
• Sputtered impurity transport w/ convective
  transport.
• Micro-structure (fuzz) formation of tungsten
  erosion.
• Also dust formation/transport, T retention.

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ReNeW PWI White Papers-Thrusts-Focus
Author (lead) Modeling Experiment (existing tokamaks /diagnostics. modest upgrades) Experiment (Other Facility use/upgrade) Major New Facility/ Modifications
LaBombard v v v
Leonard v v
Rognlien v
Brooks v
Stotler v
Krstic v
Strait v
Allain v v
Stangeby v v
Canik v v
Goldston v
Skinner v v
Dippolito v
Kotschenreuter v
Hassanein v v
Modeling tasks generally includes analysis of
experiments/code-data validation
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• Plasma Wall Interaction Panel-Potential Thrusts
• Modest effort 10 M ( 2 M/yr for 5 yrs
  w/follow-on)
• Modest enhanced effort in plasma/material
  interaction predictive modeling code
  validation.
• Moderate effort 40 M ( 8 M/yr for 5 yrs
  w/follow-on)
• More ambitious plasma/material interaction
  modeling increase major diagnostic increase
  modest facility use/upgrades innovative
  solution research
• High effort 50 M ? (5 yrs)
• Major increase in plasma/material interaction
  modeling, diagnostics, innovative solution
  research, major facility construction/upgrades.

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• Plasma Wall Interaction Panel-typical Modest
  Thrust
• GOAL Some increase in our predictive PWI
  modeling capability help identify workable
  surface materials, PFC designs, plasma operating
  parameters.
• Modest effort 10 M ( 2 M/yr for 5 yrs 5
  FTEs/yr increase) w/follow-on after the initial
  5 yr work.
• Modest enhanced effort in plasma/material
  interaction predictive modeling code
  validation.
• Areas Edge/SOL plasma with turbulence,
  sputtering erosion/redeposition, transient plasma
  effects on PFC,s, dust effects, RF sheath
  effects. Analysis of present devices, ITER,
  start of PWI DEMO analysis. Code/data validation
  efforts.
• We are on a steep portion of the learning
  curve. Thrust 1 would permit highly
  cost-effective enhancement to the existing
  highly-underfunded modeling/computation
  capability, but still leaving major gaps.
• Potentially includes small increases in
  experimental capability, e.g., addition of
  low-cost diagnostics.
• This (and all PWI research thrusts) would
  interact with thrusts/efforts to increase
  operating time, new device construction,
  supercomputer applications (e.g., Fusion
  Simulation Project), transient plasma control,
  core plasma theory/modeling, and similar relevant
  areas.

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• Plasma Wall Interaction Panel-potential Moderate
  Thrust
• GOAL Significant Increase in our predictive PWI
  modeling capability help identify workable
  materials, PFC designs, plasma operating
  parameters.
• Moderate effort 40 M ( 8 M/yr for 5 yrs 15
  FTEs) w/follow-on
• Significant plasma/material interaction modeling
  increase diagnostic increase moderate
  increased facility use/upgrades innovative
  solution research.
• Areas Includes 3-D time-dependent turbulence
  modeling, coupled (edge plasma/material
  surface/impurity transport) erosion/redeposition
  analysis, comprehensive transient analysis, dust,
  microstructural surface response, etc.
• Analysis of US devices (CMOD, NSTX, DIII-D,)
  JET, and selected other tokamaks, plasma
  simulators (PISCES, plasma guns, etc.), DEMO.
• includes moderate increases in experimental
  capability, e.g., addition of key diagnostics,
  increased operating time, but does not include
  major facility construction or major upgrades

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• Plasma Wall Interaction Panel- potential High
  Thrust
• GOAL Major increase in our predictive PWI
  modeling cabability Identify workable materials,
  PFC designs, plasma operating parameters.
• High effort 50 M ? (5 yrs) 15 FTEs/yr
  increase (note staff availability is a
  rate-limiting step).
• Includes Thrust-2 modeling goals
• Major increases in experimental capability,
  including diagnostics, operating time, new test
  facilities (e.g., lab simulator tokamak).

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Some high-leverage plasma/wall interaction
research implications

• ITER
• Keep beryllium coated wall?
• Or, dump Be, use bare wall or tungsten coated
  wall.
• Plan for existing plasma reference parameters
  (beta, confinement, Te, etc.)?
• Or, plan for reduced operation, due to
  transient PFC effects limitations.
• And/or, use innovative design solutions.
• DEMO
• Aggressively plan for liquid metal divertor RD?
• Plan for innovative solution RD.
• Have reasonable confidence that PWI issues can be
  solved?
• Or, determine that PWI is probably
  unsolvable-abandon tokamak approach ( e.g.,
  plan for fast breeder reactors).