FDF: PWI issues and research opportunities
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FDF: PWI issues and research opportunities. PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION. 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

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FDF: PWI issues and research opportunities

PERSISTENT SURVEILLANCE FOR

PIPELINE PROTECTION AND THREAT INTERDICTION

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
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
Simple estimate of rate at which tokamaks ‘work’ PFC materials

  • 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 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.

  • 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.



Net erosion in the divertor
Net erosion in the divertor

  • An FDF divertor plasma solution calculated by SOLPS (John Canik): ne ~ 1021m-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
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

  • D0 cx neutrals with E ~ 300 eV

top

ISP

OSP


Simple estimate for net wall erosion rates
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.


Rough estimate of net erosion rate of main walls
Rough estimate of net erosion rate of main walls

* Kukushkin B2-EIRENE calculation


13 c from 13ch 4 puffed into top of diii d mainly ended up in divertor
13C from 13CH4 puffed into top of DIII-D mainly ended up in divertor

outer strike point

inner strike point

puff location

inner wall

outer wall


When both divertors are detached in diii d there is net deposition in entire divertor
When both divertors are detached in DIII-D, there is net deposition in entire divertor

When both divertors are detached, there is net deposition everywhere in the divertor, evidently due to

mass transfer from the walls to the divertor.

Dennis Whyte


Have we been worrying about the wrong problem
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’.


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