Tokamak edge physics and plasma-surface interactions

1
Tokamak edge physics and plasma-surface
interactions

• Richard A. Pitts
• Centre de Recherches en Physique des Plasmas
• Ecole Polytechnique Fédérale de Lausanne,
  Switzerland
• Association EURATOM-Swiss Confederation

2
Mission statement for this talk
The interaction of plasma with first wall
surfaces will have a considerable impact on the
performance of fusion plasmas, the lifetime of
plasma-facing components and the retention of
tritium in next step burning plasma
experiments Progress in the ITER Physics Basis,
Chap. 4 Power and particle control, Nucl.
Fusion 47 (2007) S203-S263
3
Outline

• The scrape-off layer (SOL) and divertor
• SOL power width
• Divertor detachment
• SOL flows
• Plasma-surface interactions
• Material lifetime erosion and migration
• Tritium retention
• Dust
• Mixed materials

CAVEAT Edge plasma physics and PSI is a vast
domain. Can only scratch the surface in a single
tutorial. Work referenced throughout the talk is
listed at the end.
4
Divertor and SOL physics
5
Terminology limiters and divertors
Scrape-off layer (SOL) plasma region of open
field lines
LCFS
Upstream
Core plasma
Core plasma
Outermidplane
Separatrix
Limiter
X-point
Private plasma
Vessel walls
Outer
Inner
Divertor targets
6
e.g. Limiter and divertor phases in many JET shots
JET 62218, H-mode, Ip 3.0 MA, Bj 3.0 T
Ramp-up and ramp-down phases in ITER will be in
limited phase, 30 s long 5. Full burn divertor
phase of 400 s for the QDT 10 inductive
scenario
7
Basics SOL width, ln 1

• Any solid surface inserted into a plasma
  constitutes a very strong particle sink
• In the high tokamak B-fieldG? ltlt G
• Thin Debye sheath (lD few 10s mm thick ) forms
  at the surface ? controls flow of particles and
  energy B

Adapted from 1

• Quick and dirty estimate of ln with diffusive
  approx. for cross-field particle transport (all
  ionisation inside LCFS)G? ? nv? -D?dn/dr
  D?n/ ln? v? ? D?/ln , ln t?v? ? v ? cs
  (kT/mi)1/2 ? Then, if t? t,

e.g. L 30 m (typical of JET) TLCFS 100 eV,
cs 105 ms-1, D? 1 m2s-1 (near SOL)? ln 1.7
cm!!cf. minor radus 2.0 m for ITEREven worse
for energy see next
8
The problem with lq

• SOL width for power, lq, is also small and is an
  important parameter of the edge plasma
• As for particles, lq is determined by the ratio
  of ? to transport (e.g. cross-field ion
  conduction and parallel electron conduction i.e.
  ? (??/?)1/2 ), where ?? is anomalous
• Scalings for lq can be derived from models and
  experiments, e.g.
• 2-point analytic modelling PSOL
  power into SOL 1
• Scaling from H-mode experiments on JET 6
• ITER modelling 7 assumes lq 5 mm, JET scaling
  gives lq 3.7 mm (cf. a2.0 m)
• Very recent multi-machine scaling 8 gives lq/R
  constant
• Note also that the parallel power flux, q ?
  PSOL/lq as much as 1 GWm-2 in ITER

Stored energy scales strongly with tokamak major
radius, W ? R4 9But power deposition area in
the divertor ? Rlq only (3.0 m2 in ITER) Bottom
line is that despite its increased physical size,
ITER will concentrate more power into a narrower
channel at the plasma edge than todays devices.
The use of divertor detachment, radiation and
geometry will be used to reduce the surface power
flux densities to manageable levels, but careful
monitoring will be critical (unlike todays
expts., where it is useful)
9
Power handling ITER case (approx)
lq 5 mm
100 MW
q,u 500 MWm-2

• Max. steady-state power flux density permitted at
  ITER divertor targets q? ? 10 MWm-2
• Magnetic and divertor geometry alone cannot
  reduce the power to tolerable levels
• Most of the parallel power flux must be prevented
  from reaching the plates? divertor detachment
  and high radiative loss

CORE PLASMA
Magnetic flux expansion (Bq/B)u/(Bq/B)t 4 for
ITER outer divertor ? low field line angles at
strike points (3º) Target tilting in
poloidal plane (a 25º for ITER outer target)
SOL
a
per target
(adapted from 10)
q? 25 MWm-2 per target if no radiative (or
other) dissipation
10
The route to detachment (1)
Mean free paths for particle collisions are long
SOL collisionality is low Power flow to
surface largely controlled by target sheath g
sheath heat transmission coefficientepot
potential energy per incident ion
11
The route to detachment (2)
At sufficiently low Tt, (lt 5 eV), neutral
ionisation rate lt ion-neutral friction processes
(CX, elastic scattering). Momentum transferred
from ions to dense cloud of neutrals in front of
the plate (recycle region) ? begins to reduce nt,
?p ? 0 and plasma pressure falls across recycle
region.Once Tt 1-2 eV (and if nt high enough),
volume recombination locally extinguishes
plasma, reducing target power flux
Adapted from 2
Detachment seen experimentally in many devices,
but complex volumetric process and relative
importance of ion-momentum friction vs.
recombination still unclear. X-point geometry ?
long connection lengths ? high residence times in
low Te plasma ? efficient radiative loss
favouring power reductions where q is highest
(i.e. on flux surfaces near separatrix).
C-Mod, B. Labombard, et al., 11
12
Full detachment is a problem

• Detachment which is too strong (particle flux
  reduced across the whole target) is often
  associated with zones of high radiation in the
  X-point region and confined plasma (MARFE)
• MARFE formation can drive a transition from H to
  L-mode (H-mode density limit) or disruption
• MARFE physics still not well modelled

JET, A. Huber, et al. 12
Limit detachment to regions of highest power flux
(where it is needed most).Maintain remainder of
SOL in high recycling (attached)A few ways to
arrange that this happens more readily
Divertor closure
Target orientation
Impurity seeding
13
Divertor closure
Increasing closure
JET, R. D. Monk, et al. 13

• Increased closure significantly improves divertor
  neutral pressure ? increased neutral density
  (nn), promoting earlier detachment
• Closing bypass leaks important for increasing
  nn
• Divertor closure also promotes helium compression
  and exhaust very important for ITER and reactors

14
Target orientation
AUG, A. Kallenbach, et al. 14

• Parallel heat fluxes significantly reduced for
  vertical cf. horizontal targets
• Underlying effect is preferential reflection of
  recycled deuterium neutrals towards the separatrix

Separatrix
Increased ionisation near sep.
Higher nt, lower Tt
Higher CX losses
Pressure loss ? q ?
15
Impurity seeding
JET, G. F. Matthews et al. 16
DIII-D, C. J. Lasnier, et al. 15
Unfuelled
D2 puff92 torrls-1 for 1.8 s
Strong D2 puff
Ne puff12 torrls-1 for 0.1 s
Strong D2 N2 puff
Strong impurity seeding also reduces ELM size but
high price can be paid in confinement
16
ITER divertor achieves partial detachment
Inner strike pt.
Outer strike pt.
Power load (Wm-2)
ITER Divertor DDD 17, Case 489 (SOLPS5 runs by A.
Kukushkin)
Deep V-shaped divertor, vertical, inclined
targetsDome separating inner and outer targets
also helpful for diagnostics, neutron shielding
and reducing neutral reflux to the core
Kirschner et al. 17
17
Divertor exhaust
Apart from power handling, primary function of
divertor is to deal with He from fusion reactions
? compress D, T, and He exhaust as much as
possible for efficient pumping (and therefore
also good density control).
Critical criterion for an ITER burning plasma is
that He is removed fast enough such that
is satisfied. is the global helium
particle residence time a function of tp, the
He neutral density in the divertor and the
pumping speed (conductance) 18. Helium
enrichment is the ratio of He concentration
in the divertor compared to the main plasma.
To cryopumps
18
Plasma-surface interactions
19
Upscale to ITER is a big step
Extracted from G. F. Matthews et al. 19
Parameter JET MkIIGB(1999-2001) ITER ITER
Integral time in diverted phase 14 hours 14 hours 0.1 hours
Number of pulses 5748 5748 1
Energy Input 220 GJ 220 GJ 60 GJ
Average power 4.5 MW 4.5 MW 150 MW
Divertor ion fluence 1.8x1027 1.8x1027 6x1027
Code calculation
1 ITER pulse 0.5 JET years energy input
1 ITER pulse 6 JET years divertor fluence
20
ITER materials choices

• Be for the first wall
• Low T-retention
• Low Z
• Good oxygen getter

• For H and part of D phase C for the targets
• Low Z
• Does not melt
• Excellent radiator
• W for the dome/baffles
• High Yphys threshold

Beryllium
Driven by the need for operational flexibility

• For D and DT phases
• Be wall, all-W divertor

To avoid problem of T-retention
What are the issues associated with
plasma-surface interactions?
21
Critical issues
Long term tritium retention
Short and long range material migration
Material mixing
All strongly interlinked
Steady state erosion
Transient erosion(ELMs, disruptions)
Material lifetime
Redeposition
22
Impurity migration
Transport
Erosion
Deposition

Migration
Re-erosion
23
Erosion Physical and chemical sputtering
Chemical (carbon)
Physical
Adapted from Eckstein et al. 20
Roth et al. 21
ITER divertor flux
D impact

• Energy threshold ? higher for higher Z substrate
• Much higher yields for high Z projectiles
  important if using impurity seed gases

• No threshold
• Dependent on bombarding energy, flux and surface
  temperature

Current steady state divertor target erosion
rates (ERO modelling) due to Yphys and Ychem
estimated at 0.4 - 2 nms-1 for ITER 17
24
Erosion transients, e.g. ELMs on the divertor
Important factor is max. DTsurf due to arrival of
short heat pulse (duration, t) f fraction
of Wth lost during transientAdiv, divertor
wetted arear,k,Cdensity,conductivity,heat
capacity ELM energy losses must stay below
melting/sublimation/evaporation limit to avoid
fast erosion (e.g. melt later loss)
? Important to measure Tmax
Time (ms)
Federici et al. 22
If Dtransient few mm and target thickness cm
? lifetime 104 events103 Type I ELMs/discharge
? lifetime 10 ITER pulses!!
BUT tests on ITER target mock-ups with realistic
energy fluxes show that damage threshold 2x
lower than for ideal materials (crack formation
on W, PAN fibre erosion in CFC) 23,24 ? Need to
keep ELM energy flux ? 0.5 MJm-2 for W and CFC
25
Erosion ELM size must be mitigated in ITER
Materials tests (QSPA and MK-200UG plasma gun
facilities in RF) show that CFC PAN fibre erosion
and W surface cracking do not occur for energy
densities ? 0.5 MJm-2 for 500 ms, triangular
pulse lengths (250 ms up and down). How does this
convert to an ITER tolerable ELM size 25?
What do we know about ELMs at the divertor
target?1) Rise time to peak Tsurf t 26
(t cs/L 250 ms for ITER)2) Toroidal
peaking factor 1 273) lq,ELM lq,inter-ELM
5 mm (ITER)4) A?,IN 1.4 m2, A?,OUT 1.9
m25) No ELM energy to main walls6) Strong
in-out asymmetry in ELM power loading (not
understood yet)
DWELM qELM ? A?,in ? (1 Eout/Ein) 0.5 MJ/m2
? 1.4 m2 ? 1.5 1 MJ This is only 0.3 of
plasma stored energy for ITER QDT 10 baseline
scenario!Natural Type I ELMs this small do not
existAnd note that ELMs have a size
distribution!? Mitigation techniques will be
required in ITER (e.g. pellet pacing (see talk by
P.Lang), active ergodisation coils 30, 31)
26
Transport creates and moves impurities
Ions
Cross-field transport turbulent driven ion
fluxes can extend into far SOL? recycled
neutrals? direct impurity releaseELMs can also
reach first walls
Eroded Impurity ions leak out of the divertor
(?Ti forces)
SOL and divertor ion fluid flows can entrain
impurities
Neutrals

• From divertor plasma leakage, gas puffs, bypass
  leaks ? low energy CX fluxes ? wall sputtering
• Lower fluxes of energetic D0 from deeper in the
  core plasma
• A problem for first mirrors ? see talk by A. H.
  Donne

27
Migration balance example from JET

• Make balance for period 1999-2001 with MarkIIGB
  divertor 14 hours plasma in diverted phase
  (50400 s, 5748 shots)
• Use spectroscopy and modelling to estimate main
  chamber sources

• Post mortem surface analysis
• Deposition almost all at inner divertor
• Surface layers are Be rich ? C chemically eroded
  and migrates, Be stays put
• Outer divertor region of net erosion or
  balanced erosion/redeposition BUT mostly
  attached conditions (not like ITER)

20g Be (BeII)
450g C (CIII)
Main chamber source of net erosion
250 kg/year if JET operated full time!Carbon
migrates to remote locations forming D-rich soft
layers (high T-retention)
22g Be
Pitts et al. 4
Matthews et al. 32
Strachan et al. 33
Likonen et al. 34
28
Tritium retention (1)

• One of the most challenging operational issues
  for burning plasmas
• If carbon present, complex interplay between
  erosion ? hydrocarbons ? dissociation/ionisation
  ? transport ? re-deposition ? migration to
  remote areas with high sticking coefficients and
  retention in co-deposits
• Carbon traps D, T very efficiently
• D/C ratio can be in the range 0.4 ? gt 1
  depending on the type of re-deposited layer
• Retention very hard to characterise in todays
  mostly carbon dominated devices
• Dependent on materials, Tsurf, geometry
  (limiter/divertor), operating scenarios (H-mode,
  L-mode, low/high dens.)

Reported measurements range from 3-50 retention
35! e.g. on JET, 3 obtained from long term,
post mortem surface analysis, 10-20 from gas
balance.
29
Tritium retention (2)
C targets, Tsurf 800ºC, chemicalphysical
sputtering

• A 400 s QDT 10 ITER discharge will require 50
  g of T fuelling(cf. 0.01-0.2 g in todays
  tokamaks)
• Working guideline for max. in-vess. mobilisable T
  in ITER 1kg 18,36
• World supply of T is also limited
• Must avoid build-up in inaccessible locations
• Predicting the expected retention in ITER is
  notoriously difficult

2g/discharge
0.02g/discharge
Roth et al. 37
ITER target 18 is a retention level of 0.05
g/discharge ? 7000 shots before major shutdown
for T-removal

• Very recent estimates (ERO code 2007 including Be
  main chamber influx) show that the in-vessel
  limit could be reached after only 140 shots 17
• Modelling does not yet contain effects of
  transients (ELMs disruptions)!
• No account taken for trapping in tile gaps

Accurate measurement of T-retention and the
development of efficient T-removal methods will
be critical for the success of ITER
30
Dust

• Dust is seen in all tokamaks, especially with C
  walls
• Not generally a concern in todays devices
• But is potentially very important in ITER
• As an inventory for trapped tritium in areas
  difficult to access
• As an explosive safety hasard water leak ? hot
  surfaces ? steam ? hydrogen (by oxidation) ?
  possible explosion if enough air also present
  38
• As a radiological or toxic hasard (activation
  products of W, tritium contained in Be, C dust,
  toxicity of Be dust)

TCV floor viewing IR camera during disruption,
33448 (J. Marki, R. A. Pitts
Carbon dust collected from tokamaks after
operation periods is usually micron sized. Formed
from flaking and degradation of deposited films,
unipolar arcing, brittle fracture (e.g. due to
transients) etc.
No real idea yet how much dust ITER will
generate, where it will go or how to get it out ?
a big effort needed to improve this situation
DIII-D floor viewing DiMES TV with near IR
filter. 2nd shot in 2007 after dirty vent,
127331. Courtesy of D. L. Rudakov W. P. West
39
31
Mixed Materials
No fusion device operating today contains the
material mix currently planned for the ITER first
wall and divertor Be, W, C. Cross contamination
of the material surfaces will be unavoidable.
This is likely to have several consequences 18

• Formation of metallic carbides ? diffusion of C
  into bulk material at high temperatures
• Formation of Be-W alloys ? melting point can be
  reduced by as much as 2000ºC

Material property changes due to mixing

• Retention of H in BeO can be as high as in C
• Retention in W can be increased by C or oxide
  layers but is very low in pure W or Be
• Very complex difficult to predict yet for ITER

Effect on H-isotope retention

• Can both increase and decrease erosion!
• Heavy ions (e.g. BeZ, CZ) on C, W ? increased
  phys. sputt. but surface coverage (e.g. Be on C)
  reduces chemical sputtering.

Effect on material erosion
Preparations underway at JET to test a Be/W
Be/W/C wall mix from 2011 40
32
References (1)
A 40 min. talk can only hope to scratch the
surface of such a vast field. Some good
reference sources covering aspects of the
material shown in this presentation are the
following
1 The plasma boundary of magnetic fusion
devices, P. C. Stangeby, IoP Publishing Ltd,
Bristol, 2000 2 Experimental divertor
physics, C. S. Pitcher and P. C. Stangeby,
Plasma Phys. Control. Fusion 39 (1997) 779 3
Plasma-material interactions in current tokamaks
and their implications for next step fusion
reactors, G. Federici et al., Nucl. Fusion 39
(1997) 79 4 Material erosion and migration in
tokamaks, R. A. Pitts et al., Plasma Phys.
Control. Fusion 47 (205) B303
A number of additional papers have been used to
prepare the slides in this presentation. They are
listed below in order of appearance in the talk.
5 Simulations of ITER start-up and assessment
of initial power loads, G. Federici et al., J.
Nucl. Mater. 363-365 (2007) 346 6 Boundary
plasma energy transport in JET ELMy H-modes, W.
Fundamenski and W. Sailer, Nucl. Fusion 44 (2003)
20 7 Scaling laws for edge plasma parameters
in ITER from two-dimensional edge modelling, A.
Kukushkin et al., Nucl. Fusion 43 (2003) 716 8
Plasma-surface interaction, scrape-off layer and
divertor physics implications for ITER, B.
Lipschultz et al., Nucl. Fusion 47 (2007)
1189 9 Steady state and transient power
handling in JET, G. F. Matthews et al., Nucl.
Fusion 43 (2003) 999 10 The plasma-wall
interaction region a key low temperature plasma
for controlled fusion, G. F. Counsell, Plasma
Sources Sci. Technol. 11 (2002) A80 11
Experimental investigation of transport
phenomena in the scrape-off layer and divertor,
B. LaBombard et al., J. Nucl. Mater. 241-243
(1997) 149 12 Improved radiation measurements
on JET first results from an upgraded bolometer
system, A. Huber et al., J. Nucl. Mater.
363-365 (2007) 365 13 Recent results from
divertor and scrape-off layer studies on JET, R.
D. Monk et al., Nucl. Fusion 39 (1999) 1751 14
Scrape-off layer radiation and heat load to the
ASDEX Upgrade LYRA divertor, A. Kallenbach et
al., Nucl. Fusion 39 (1999) 901
33
References (2)
15 Study of target plate heat load in diverted
DIII-D tokamak discharges, C. Lasnier et al.,
Nucl. Fusion 38 (1998) 1225 16 Studies in JET
divertors of varied geometry II Impurity seeded
plasmas, G. F. Matthews et al., Nucl. Fusion 39
(1999) 19 17 Modelling of tritium retention
and target lifetime of the ITER divertor using
the ERO code, A. Kirschner et al., J. Nucl.
Mater. 363-365 (2007) 91 18 ITER Physics
basis Chapter 4, power and particle control,
Nucl. Fusion 39 (1999) 2391 19 Material
migration in JET, G. F. Matthews et al., Proc.
30th EPS Conf. on Control. Fusion and Plasma
Physics (St. Petersburg, 2003) 27A (ECA)
P-3.198 20 W. Eckstein et al., IPP Garching
report number 9/82 (1993) 21 Flux dependence
of carbon chemical erosion by deuterium ions, J.
Roth et al., Nucl. Fusion 44 (2004) L21 22
Assessment of the erosion of the ITER divertor
targets during Type I ELMs, G. Federici et al.,
Plasma Phys. Control. Fusion 45 (2003) 1523 23
Transient energy fluxes in tokamaks Physical
processes and consequences for next step
devices, A. Loarte et al., 34th EPS Conf. on
Control. Fusion and Plasma Physics (Warsaw,
2007) 24 Effect of ELMs on ITER divertor
armour materials, A. Zhitlukhin et al., J. Nucl.
Mater. 363-365 (2007) 301 25 A. Loarte et al.,
to be presented at 22nd IAEA Fusion Energy
Conference, Geneva, Switzerland, October 2008
26 Power deposition onto plasma-facing
components in poloidal divertor tokamaks during
Type I ELMs and disruptions, T. Eich et al., J.
Nucl. Mat. 337-339 (2005) 669 27
Characteristics of Type I ELM energy and
particle losses in existing devices and their
extrapolation to ITER, A. Loarte et al., Plasma
Phys. Control. Fusion 45 (2003) 1549 28
Divertor power deposition and target current
asymmetries during Type I ELMs in ASDEX Upgrade
and JET, T. Eich et al., J. Nucl. Mat. 363-365
(2007) 989 29 ELM transport in the JET
Scrape-off layer, R. A. Pitts et al., Nucl.
Fusion 47 (2007) 1437 30 Suppression of large
Edge-Localised Modes in high confinement DIII-D
plasmas with a stochastic magnetic boundary, T.
E. Evans et al., Phys. Rev. Lett. 92 (2004)
235003 31 Active control of Type I
Edge-Localised Modes with n 1 perturbation
fields in the JET tokamak, Y. Liang et al.,
Phys. Rev. Lett. 98 (2007) 265004
34
References (3)
32 Material migration in divertor tokamaks,
G. F. Matthews et al., J. Nucl. Mater. 337-339
(2005) 1 33 JET carbon screening experiments
using methane gas puffing and its relation to
intrinsic carbon impurities, J. D. Strachan et
al., Nucl. Fusion 43 (2003) 922 34 Beryllium
accumulation at the inner divertor of JET, J.
Likonen et al., J. Nucl. Mater. 337-339 (2005)
60 35 Gas balance and fuel retention in fusion
devices, T. Loarer et al., Nucl. Fusion 47
(2007) 1112 36 Overview of goals and
performance of ITER and strategy for plasma-wall
interaction investigation, M. Shimada et al., J.
Nucl. Mater. 337-339 (2005) 808 37 Flux
dependence of carbon erosion and implications for
ITER, J. Roth et al., J. Nucl. Mater. 337-339
(2005) 970 38 The safety implications of
tokamak dust size and surface area, K. A.
McCarthy et al., Fus. Eng. Design 42 (1998)
45 39 Observations of Dust in DIII-D Divertor
and SOL, D. L. Rudakov et al., 1st Workshop on
Dust in Fusion Plasmas, 8-10 July 2007,
Warsaw, Poland 40 An ITER-like wall for JET,
J. Pamela et al., J. Nucl. Mater. 363-365 (2007) 1
35
Reserve slides
36
Edge Diagnostics on ITER will be critical
Courtesy of A. Kukushkin
Te
ne
Prad
Wall temperature and visible imageMain chamber
gas pressure and gas compositionSOL neutral
density (D/T)Impurity influxesSOL ne, Te
profiles (but challenging)
ne, Te, Prad, position of ionisation front, nHe,
nD/nT, impurity and D, T influxes
Cryopump inlet composition (H/D/T/He, CxHy)
Target plate heat and particle fluxes, Te,
surface temperature, erosion rate
Neutral gas pressure
Dust accumulation?
See talk by A. H. Donne