PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION FDF: PWI issues and research opportunities Peter Stangeby University of Toronto Presented at the ReNew Theme III workshop; Taming the Plasma Material Interface UCLA, March 4-6, 2009 UNIVERSITY OF TORONTO Institute for Aerospace Studies

Reactors will make their own PFCs to interact with PWI in present devices usually does little to the PFC material. The plasma is essentially still interacting with the material that was installed. In reactors, however, the PWI will strongly ‘work’ the PFC material, actually creating the wall material that the plasma reacts with. This situation will be so different from what we see today in fusion devices that we have little reliable idea of the consequences. Successful development of fusion power therefore requires that study of PWI on PWI-created PFCs begin as early as possible. This requires facilities that create far more intense PWI than do present devices.

Simple estimate of rate at which tokamaks ‘work’ PFC materials Assume P rad = 75% P heat, thus 0.25P heat = γkT s φ s, where γ = sheath heat transmission coefficient = 7; T s = plasma average temperature in contact with surfaces = 10 eV assumed here; φ s = 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 C surface temperature of 1000C. Y eff (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 P rad, here 3% C3+ in the target ion flux (~ same effect for N3+). Y eff (W) = 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.

Rate at which PFC materials are worked in various devices deviceP heat [MW] annual run time [s/year] fluence of D/T ions to surfaces [TC/yr] * beryllium circulation rate [kg/yr] boron circulation rate [kg/yr] carbon circulation rate [kg/yr] tungsten circulation rate [kg/yr] DIII-D JT 60SA EAST ITER FDF Reactor x ,00039,00023,00035,000 *tera-coulombs = 6.25x10 30 DT-ions

Net erosion in the divertor An FDF divertor plasma solution calculated by SOLPS (John Canik): n e ~ m -3, T ~ 10 eV at outer strike point. Ionization mfp of physically sputtered Be, B, C ~ 0.3mm ~ ion larmor radius. Thus probability of prompt local redeposition ~ 1. Thus net erosion << gross erosion expected. DIVIMP code (David Elder) applied to SOLPS plasma solution, finds carbon net erosion << gross erosion.

ITER divertor and wall fluxes calculated using B2-EIRENE (Kukushkin) 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 D 0 cx neutrals with E ~ 300 eV ISP OSP top

Simple estimate for net wall erosion rates Assume physical sputtering for cx neutral tritons only. Yields for E cx = 300 eV T (Eckstein 2002). Normal incidence yields doubled to account for surface roughness: for (Be, B, C, W), Y cx = (0.083, 0.056, 0.035, ). No sputtering included for D 0, He 0, low-Z neutral or self-neutral and no sputtering included for ion-wall interaction. Assumes P cx = 0.05 P heat (~Kukushkin for ITER), thus 0.025P heat = E cx φ cx and gross erosion rate = Y cx φ cx ~ net erosion rate for main wall.

Rough estimate of net erosion rate of main walls For 300 eV T o cx sputtering of walls P heat [M W] annual run time [s/year] beryllium net wall erosion rate [kg/yr] boron net wall erosion rate [kg/yr] carbon net wall erosion rate [kg/yr] tungsten net wall erosion rate [kg/yr] DIII-D JT-60SA EAST ITER [29 * ]6444 [53 * ]92 [41 * ] FDF Reactor4002.5x * Kukushkin B2-EIRENE calculation

13 C from 13CH 4 puffed into top of DIII-D mainly ended up in divertor puff location inner strike point outer strike point inner wall outer wall

When both divertors are detached in DIII-D, there is net deposition in entire divertor Dennis Whyte When both divertors are detached, there is net deposition everywhere in the divertor, evidently due to mass transfer from the walls to the divertor.

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 net deposition 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 be ‘flow thru’ – or at least ‘flow in’.