Plasma Wall Interactions (PWI) Panel Introduction J.N. Brooks 1 and the ReNeW Theme III PWI-Panel 1 Purdue University ReNeW Meeting, UCLA, March 4-6, 2009

J.N Brooks, ReNeW 3/5/09 2 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)

J.N Brooks, ReNeW 3/5/09 3 ReNeW Meeting Inputs: PWI PWI Conference Calls White Papers- ~40 Theme III, ~ 15 PWI Inputs from community

J.N Brooks, ReNeW 3/5/09 4 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)

J.N Brooks, ReNeW 3/5/09 5 Theme III White Papers

J.N Brooks, ReNeW 3/5/09 6 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

J.N Brooks, ReNeW 3/5/09 7 Fusion plasma facing material requirements Heat flux –~ 10 MW/m 2 peak (ITER, on divertor), normal operation – GW/m 2 peak, w/plasma transients –~100 MW (ITER) 600 MW (commercial reactor) total surface heat load Particle flux – D-T: m -2 s eV –He +2 : m -2 s eV –O +k : ~0.1% of D-T Neutron flux –~ 0.5 MW/m 2 (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.

J.N Brooks, ReNeW 3/5/09 8 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

9 Be-W interaction can lead to extreme failure (PISCES crucibles) Intact W wall (97%W, 3%O) Inner wall coating (4% W, 95% Be, 1%O) Be 22 W? Crucible failure zone (9% W, 70% Be, 14% C, 7% O) Be 12 W?

J.N Brooks, ReNeW 3/5/09 10 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

J.N Brooks, ReNeW 3/5/09 11 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.

J.N Brooks, ReNeW 3/5/09 12 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.

J.N Brooks, ReNeW 3/5/09 13 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.

J.N Brooks, ReNeW 3/5/09 14 R. Goldston and the

J.N Brooks, ReNeW 3/5/09 15 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

J.N Brooks, ReNeW 3/5/09 16 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.

J.N Brooks, ReNeW 3/5/09 17 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)

J.N Brooks, ReNeW 3/5/09 18 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}

J.N Brooks, ReNeW 3/5/09 19 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) [2] A. Hassanein et al., PSI-18 (2008), J. Nuc. Mat. to be pub.

J.N Brooks, ReNeW 3/5/09 20 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.

J.N Brooks, ReNeW 3/5/09 21 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.

J.N Brooks, ReNeW 3/5/09 22 Author (lead)Modeling*Experiment (existing tokamaks /diagnostics. modest upgrades) Experiment (Other Facility use/upgrade) Major New Facility/ Modifications LaBombard Leonard Rognlien Brooks Stotler Krstic Strait Allain Stangeby Canik Goldston Skinner Dippolito Kotschenreuter Hassanein ReNeW PWI White Papers-Thrusts-Focus * Modeling tasks generally includes analysis of experiments/code-data validation

J.N Brooks, ReNeW 3/5/09 23 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.

J.N Brooks, ReNeW 3/5/09 24 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.

J.N Brooks, ReNeW 3/5/09 25 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

J.N Brooks, ReNeW 3/5/09 26 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).

J.N Brooks, ReNeW 3/5/09 27 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 R&D? Plan for innovative solution R&D. 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).