PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION

1. Carbon as a flow-through, consumable PFC material ARIES Town Meeting on Edge Plasma Physics and Plasma Material Interactions in the Fusion Power Plant Regime20-21 May 2010UC San Diego PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION Peter Stangeby University of Toronto UNIVERSITY OF TORONTO Institute for Aerospace Studies

2. Some designs for DT devices use tungsten PFCs.W suffers from “foam” formation. Tungsten experiences a surface roughening effect due to He bombardment at energies below the threshold for physical sputtering (simultaneous D + T + 5-10% He, as will occur in burning plasmas), creating a surface “foam” (nanoscopic morphology) [Nishijima, Baldwin, Doerner]. “Such structures are potentially large sources of high-Z dust, could harbour significant levels of retained hydrogen isotopes in the plethora of helium nano-bubbles/voids within nano-structures and may have a dramatic influence on surface-thermal properties of W plasma facing components.” [Baldwin and Doerner]. The effect becomes significant over ~1000K and could be the most serious issue for W in DEMO [Konishi and Ueda].

3. Low-Z coatings are a potential solution to the W ‘foam’ problem • A solution to the problem has been proposed [Baldwin, Doerner, Nishijima, Tokunaga, Ueda ]: coating the W with a low-Z material such as Be, B or C. • Coating W with a low-Z material could also solve the problem of unacceptably high concentrations of W in the confined plasma. • The low-Z coatings would be consumed and replenished continuously and so neutron damage may not be an issue. • Continuous refurbishment of the plasma-facing surfaces with coatings would also solve the basic problem of loss of structure in high duty cycle devices.

4. Very little tritium retention in low-Z coatings at high temperature • Reactors: hot walls, ~1000C, for thermal efficiency • At high temperatures there is little tritium retention in any material, including C, where H/C < 0.003 for T > 1000C • In hot wall devices: little T retention in low-Z co-deposits, a serious issue for cold (200-300C) wall devices like ITER • In hot wall devices: T permeation rather than retention • Low-Z coatings may also act as a permeation barrier. But might increase retention. Research needed.

5. Reactors will make their own PFCs to interact with • PWI in present devices usually does little to the PFC material. Plasma is still interacting the installed material. • In reactors, PWI will strongly ‘work’ the PFC material, (re-) creating the wall material the plasma reacts with. • This situation will be so different from what we see today that we may have little reliable idea of the consequences.

6. No chemical sputtering or radiation enhanced sputtering (RES) of carbon at 1000C • Chemical sputtering → 0 for T > ~ 700C. [Toronto, JNM 255 (1998) 153] • RES ~ 0 until T > 1100C [Tore Supra, PPCF 40 (1998) 1335]. or RES ~ 0 until T > 1900C [TEXTOR, JNM 220-222 (1995) 467].

7. Rough estimate of rate at which tokamaks ‘work’ PFC materials. • The material circulation rate = gross erosion rate = rate at which material is worked is not to be confused with the net erosion rate, which is the required (external) refurbishment rate. • Assume Prad = 75% Pheat 0.25Pheat = γkTsφs γ = 7, sheath heat transmission coefficient Ts = 10 eV, plasma average temperature at surfaces φs = total D/T-ion flux to all surfaces, targets and walls Yeff (Be/B/C/W) = 0.021/0.0097/0.005/0.0005 UNIVERSITY OF TORONTO Institute for Aerospace Studies 3

8. Gross erosion ratesin present and future devices *tera-coulombs = 6.25x1030 ions. **JET measured ~5x1021 C/s: Strachan, NF 43 (2003) 922. UNIVERSITY OF TORONTO Institute for Aerospace Studies 4

9. Net erosion in the divertor • ITER (Kukushkin, B2/EIRENE) and FDF (Canik, SOLPS) divertor solutions give ne ~ 1021m-3, T ~ 10 eV near outer strike point. • For ne = 1021m-3, T = 10 eV, ionization penetration of physically sputtered C-atoms ~ 0.04mm < DT-ion larmor radius ~ magnetic pre-sheath thickness. Thus probability of prompt local redeposition ~ 1. Thus net erosion << gross erosion expected. Same or more so for W.

10. ITER divertor and wall fluxes calculated using B2-EIRENE (Kukushkin) fluxes • The wall, however, is in a quite • different situation than the • divertor. • Impurity neutrals sputtered from main walls by cx neutrals and by far-periphery ions enter a much weaker plasma, where ionization may occur far from surface. Thus ~no prompt local redeposition and so net erosion ~ gross erosion • D0 cx neutrals with E ~ 300 eV top ISP OSP energy

11. Simple estimate for net wall erosion rates • Assume physical sputtering for cx neutral tritons only. Yields for Ecx = 300 eV T (Eckstein 2002). • Normal incidence yields doubled to account for surface roughness: for (Be, B, C, W), Ycx = (0.083, 0.056, 0.035, 0.0024). • No sputtering included for D0, He0, low-Z neutral or self-neutral and no sputtering included for ion-wall interaction. • Assumes Pcx = 0.05 Pheat (~Kukushkin for ITER), thus 0.025Pheat = Ecxφcx and gross erosion rate = Ycx φcx ~ net erosion rate for main wall.

12. Rough estimate of net erosion rate of main walls * A Kukushkin. B2-EIRENE calculation. ** K Lackner. EU EFP Workshop on W. Velence, Hungary, Dec 7-9 2009.

13. K. Lackner: W use in DEMO Expected first wall material ablation (due to CX sputtering) W erosion in 5 yrs: 1 mm – but 25 t total ! EU EFP Workshop on W. Velence, Hungary, Dec 7-9 2009.

14. When both divertors are detached in DIII-D, there is net deposition in the entire divertor DG Whyte. DiMES measurements. Evidently due to material migration from the walls to the divertor. UNIVERSITY OF TORONTO Institute for Aerospace Studies 5

15. Have we been worrying about the wrong problem? • We have been greatly concerned about the problem of net erosion at the strike points. • It may be, however, that for high power, high density plasmas, the entire divertor may be in netdeposition due to mass transfer from the walls. • Wall erosion itself may be tolerable if not too localized. • The problem, however, will be how to clear the slag out of the divertor to avoid disrupting the plasma. All PFC materials may have to be ‘flow thru’.

16. Unique advantage of C as low-Z coating: unwanted deposits easily removed by O2-baking • Accumulation of thick deposits inevitable • Even if containing no tritium -> other problems: dust, UFO-disruptions, etc • So, it is just as essential to be able to remove in situ the low-Z material as to be able to create in situ the coating in the first place. • Carbon unique: refractory, yet a gas for in & out as CH4 & CO2 • Since ~no T then ~no DTO

17. Estimates of net erosion rates and tritium codeposition rates for a reactor • Net erosion (previous table): ~3700 kgC/yr. • O2-baking at >700C is extremely rapid, such that even a graphite substrate, if present, would be removed quickly. • At 1000C, T/C < ~ 0.0015 in carbon codeposits, thus T-throughput via codeposition ~ 1.4 kg T/yr.

18. Reactivity of carbon with O2is very high for T > 700C. 2 gmC/m2/s removed in 1 atm air • 3.7 tonsC/yr net deposits • Assume C deposited on 100 m2. • Thus time to remove: • once/yr: 6 hrs • once/mo: 30 min • once/wk: 6 min • once/day: 1 min • W substrate, if exposed, • releases volatile oxides • for T > 1250C. • SiC substrate, if exposed, • degrades for T >1400C. pure graphite codeposits pure graphite

19. Conclusion • Like Li, C may be usable as a flow-through PFC material • Experiments are required to assess this possibility: - studies of very high temperature O2-baking of codeposits - assessment of collateral damage in very high temperature baking - identification of best substrate material

20. Rough estimate of rate at which tokamaks ‘work’ PFC materials. • The material circulation rate = gross erosion rate = rate at which material is worked is not to be confused with the net erosion rate, which is the required (external) refurbishment rate. • Assume Prad = 75% Pheat, thus 0.25Pheat = γkTsφs , where γ = sheath heat transmission coefficient = 7; Ts = plasma average temperature in contact with surfaces = 10 eV assumed here; φs = total D/T-ion flux to all surfaces [ions/s], targets and walls. • Be, B, C sputtering: physical due to D/T-ions and self-sputtering. Carbon chemical sputtering and RES assumed to be not significant at assumed reactor-like C surface temperature of ~1000C. • Yeff (Be/B/C) = 0.021/0.0097/0.005 (Eckstein 2002 yields for maxwellian ions plus a 3kT-sheath). • W sputtering is due to (i) self-sputtering, and (ii) sputtering by a low-Z additive required to increase Prad, here 3% C3+ in the target ion flux (~ same effect for N3+). Yeff (W) = 0.0005. UNIVERSITY OF TORONTO Institute for Aerospace Studies 3