Power and particle exhaust :

1
Power and particle exhaust introduction,
recent progress on JET and implications for
ITER W.Fundamenski Culham Centre for Fusion
Energy (Euratom/CCFE Fusion Association) on
behalf of members of the EFDA-JET Exhaust Task
Force including T. Eich, S. Brezinsek, G.
Maddison, K. McCormick, C. Giroud, G.Arnoux, S.
Devaux, M. Jakubowski, D. Tskhakaya, A. Loarte,
V. Riccardo, H. Thomsen, A. Alonso, P. Belo, M.
Beurskens, I. Coffey, R. Felton, J. Harling, A.
Huber, G. Sergienko, S. Jachmich, E. Joffrin, J.
Ongena, P. Monier-Garbet, P. Debynck, M. Stamp
and JET EFDA Contributors
PPPL Colloquium Princeton, 11 November 2009
2
W.Fundamenski UKAEA, EFDA-JET ITER Summer
School Aix-en-Provence, 23 June 2009
3
Ignition vs. Exhaust Criteria
achieve
maintain
QDT Pfus / Pheat 10
Fuelling
Particle exhaust
gas
pellets
beams
Intrinsic Z
Extrinsic Z
He ash
Heating
Power exhaust
alpha
RF
NBI
neutron
photon
plasma
Confinement
First wall design
equilibrium
mechanical
stability
thermal
current drive, disruptions, tritium, dust,
transport
nuclear
4
max QDT function(reactor design)
Ignition systems
Exhaust systems
Fuelling systems
Control system
QDT Pfus / Pheat pDT tE f(Zeff)
Heating systems
Cooling circuit
Current drive
Maximum achievable QDT for a given reactor design
is determined by exhaust limits
Cryopumps
Reactor design
PFCs
Magnetic coils
Plasma scenario
Fuelling
Particle exhaust
Heating
Power exhaust
Current profile
PFC loads
Stability transport
Confinement
Impurity influx
Edge plasma conditions
Core plasma conditions
5
Transient heat load limits in ITER
TRINITI plasma gun
CFC
40
250 us
W
ITER adopted 0.5 MJ/m2 for the maximum allowed
ELM energy load in 250 us
6
Power balance on ITER
Pfus 400 MW
Fusion
Wall area
Heating
QDT 10
Ph 40 MW
Pa 80 MW
Pn 320 MW
Alphas
Neutrons
800 m2
PSOL 60 MW
Pradcore 60 MW
Plasma
Photons (50)
?
800 m2
PELM 20 MW
Pinter-ELM 40 MW
Inter-ELM
ELMs
?
2p x 5 m x 4 mm x 10 1.2 m2
14 MW
inner
6 MW
inner
?
?
?
6 MW
outer
34 MW
2p x 6 m x 4 mm x 10 1.5 m2
outer
Outer divertor losses (CX, ES,
radiation)
25 MW
15 MW
1.5 m2
Total plasma (outer)
Steady-state design limit 10 MW/m2
0.5 MJ / m2
1.2 m2
In reality, must repeat backwards to find maximum
achievable QDT for given PFC limits
ELM frequency 20 Hz
Transient design limit 0.5 MJ/m2 in 250 us
Plasma purity (Zeff 1.7) requires high density
(fGW 0.85) and cold divertor (lt 5 eV)
Need 85 MW total radiation (70 total 50 in
core 20 in SOL), and ELM frequency above 20
Hz (ELM size 1 MJ or 1 of Wped)
7
ELM control techniques
technique not optimized
Consider the best results achieved so far
8
Ignition vs. Exhaust beta limits
9
ITER-like wall (ILW) on JET
JET ILW is a test bed for plasma operation and
material migration in ITER with the material mix
for the DT phase
ITER

• First time the Be-W combination will be used in
  a tokamak
• Carbon-free environment
• Impact on tritium retention
• Ratio of C to Be divertor influx drastically
  changed
• Absence of carbon layers (?)
• Carbon to be replaced as radiator (Ar, Ne)
• Sequential start of operation (restart, stepwise
  
• increase of power, inspections, etc.)
• New operation instructions and limitations

JET
Option 1
Reference discharges to document C and Be
levels in different configurations
Courtesy of G.F. Matthews, V. Philipps, S.
Brezinsek
10
Physical vs chemical erosion
Physical sputtering yield
Chemical erosion yield (D on C)
increases with projectile energy and mass, while
decreasing with target (PFC) material atomic mass
decreases with D ion flux and is sensitive to C
target temperature
11
ITER-like wall (ILW) on JET
Main chamber Be and W-coatings
Divertor W-coatings and bulk W outer target
GFMatthews PFMC V.Riccardo et al. PSI
conference 2008
12
Outline

• Preface
• Motivation, steady transient heat loads in ITER
  ILW
• Introduction to Plasma Exhaust
• Divertor vs Limiter SOL, edge-SOL turbulence,
  divertor regimes
• Divertor heat loads
• Divertor geometry and IR, ELM filaments, H/D/He
  comparison
• Impurity seeding
• Ne vs N2, effect on inter-ELM vs ELM, H98 Zeff
• Disruptions
• Description of DMV MGI, TQ heat loads, CQ
  runaways
• Material migration
• Be evaporation, QMBs, C-spectroscopy detachment
• Fuel retention
• Outline of gas balance experiments
• Modelling
• EDGE2D/EIRENE, ESELSOLF1D, kinetics
• Conclusions

13
Limiter vs divertor exhaust
Scrape-off layer (SOL) plasma
SOL
SOL
Edge
Edge
SOL
Core
upstream nu , Tu
Core
Limiter
LCFS
Separatrix
Vessel walls
Private plasma
Target nt , Tt
Divertor targets
14
Limiter vs divertor recycling
SOL
Limiter
C0, CxDy
D2, D0
D2, D0
Impurities, eg. C0, CxDy
lD0 few cm, lC0 1 cm, lCxDy few mm
Divertor target
Intimate contact with edge plasma
PFCs removed from edge plasma
Little recycling/cooling in the SOL results in a
hot, tenous SOL plasma
Colder, denser SOL plasma, due to local recycling
/ cooling
High erosion yields, poor pumping
Lower erosion yields, improved pumping
Strong influx of both fuel and impurity neutrals
into the edge
Fuel and impurity sources screened from the edge
by the divertor plasma
Improved plasma purity, i.e. lower Zeff
Impure edge core, i.e. high Zeff
15
Limiter vs divertor exhaust
Scrape-off layer (SOL) plasma
SOL
SOL
Edge
Edge
SOL
Core
upstream nu , Tu
Core
Limiter
LCFS
Separatrix
Vessel walls
SOL width determined by competition between
parallel and perpendicular transport
Private plasma
Target nt , Tt
Steady-state plasma loads determined largely by
Edge/SOL turbulence !
Divertor targets
16
Plasma turbulence in the Edge-SOL
Collisionality scan
Current scan
Density scan
ESEL
Wall flux 1/current
Wall flux density2
TCV
O.E.Garcia et al, PPCF 48 (2006) L1
17
JET limiter SOL profiles
Similarly, limiter SOL width decreases with
increasing plasma current (decr. L)
Physical mechanism not understood at the time !
1 MA
J.A.Tagle et al, 14th EPS 11C (1987) 662
2 MA
S.K.Erents et al, NF 28 (1988) 1209
JET
2 MA
3 MA
2 MA
3 MA
4 MA
Ip
3 MA
4 MA
5 MA
4 MA
Ip
5 MA
Ip
5 MA
18
Density profiles in the Edge-SOL
TCV
C-mod
ESEL
n
n
SOL density profile broadens with increasing
collisionality
ESEL
Such broadening observed on many tokamaks
Density fluctuations increase with radius,
approaching unity in the far-SOL
19
Radial flow profiles in the Edge-SOL
TCV
Radial plasma flux increases with collisionality
Such increase with radius observed on many
tokamaks
ESEL
Effective radial velocity is roughly constant
with radius and increases with collisionality in
the near-SOL
ESEL
20
PDFs of fluctuations in the far-SOL
TCV
Similar result for PDF of velocity fluctuations
Temporal pulse shape of density blobs reveals
leading front trailing wake
PDF of density fluctuations in the far-SOL
universal and highly intermittent
21
Interchange motion of plasma blobs
pulse shape
interchange drive
Dynamics of plasma filaments, or blobs, is
determined by charge conservation balance of
divergences of polarization, diamagnetic and
parallel currents
sheath dissipation
acceleration
O.E.Garcia et al, PoP (2006)
As collisionality increases, plasma filaments
become electrically isolated from the sheath at
the divertor target, making the interchange drive
more effective
O.E.Garcia et al, PPCF 48 (2006) L1
22
Plasma turbulence in the Edge-SOL
Mostly drift-Alfven dynamics in the edge region
Intermittent transport implies strong
fluctuations in far-SOL quantities
Mostly interchange dynamics in the SOL region
Local flux not related to local gradient!
Turbulence driven by edge pressure gradients,
which build up together with poloidal flow shear,
Mean field approximation, used in most edge
transport codes, is not accurate
damped by parallel losses and sheath dissipation
ltnTgt ? ltngtltTgt, etc.
Hence, need global edge turbulence codes
Quiescent periods interrupted by intermittent
ejection of plasma filaments
These advect mass and energy into the far-SOL,
while draining to the divertor
23
2-D Edge-SOL turbulence blobs

• Output of an ESEL simulation of SOL turbulence

Courtesy of E. Havlickova V. Naulin
24
Divertor operating regimes
Low recycling (sheath limited) nt
nu, Tt Tu, pt pu
high recycling (conduction limited) nt
nu3, Tt nu2, pt pu
Can the narrow SOL heat load profile be broadened
by a divertor buffering ?
upstream nu , Tu
Fully detached (radiation limited)
X-point MARFE
Partially detached (CX-ES limited) nt
decreases, Tt lt 5 eV, pt ltlt pu
Target nt , Tt
25
Divertor plasma detachment
A.Loarte et al, PPCF (2001)
Loss of plasma pressure and energy by CX/ES
line radiation
JET
Reduction of target plasma flux
Inner target typically detaches earlier, i.e. at
lower upstream density, than the outer target
This asymmetry is consistent with power flow into
divertor volume (ExB drifts, geometry, etc.)
Detachment of the outer divertor is needed for
steady-state load reduction
26
2-D multi-fluid detachment
log Te

• previous 2D modelling of density ramp-up
  scenario showstrong disagreement with experiment
  at transient into divertor detachmenttoo quick
  transition into full detachment,no inner/outer
  asymmetry,no real flux roll-over? transport
  model assumptions insufficient?
• there are possible candidates to revise
  transportmodel ? positive collisionality
  dependence of radial transport(reasoned by
  several experiments, eg at TCV, PPCF 2007)
• density ramp-up modeling with EDGE2D-EIRENEcan
  in principle deliver solutions to the above
• in/out asymmetry achieved, ie longer time-delay
  betweeninner and outer target detachment
• more pronounced flux-rollover
• observation of oscillations in solutions in case
  offull detachment, (on time-scale of a few Hz)?
  likely explanation for experimentally measured
  oscillations near L-mode density-limits, e.g. on
  JET. Loarte et al, PRL 1999

Courtesy of S.Wiesen
nspx
27
Thermal instability X-point MARFE
Outer target detachment typically accompanied by
an X-point MARFE
Results in substantial cooling of the edge
plasma, reduction of pedestal stored energy and
degradation of energy confinement
At higher densities transforms into an inner wall
MARFE density limit nGW Ip/a2
JET
A. Huber et al, NF (2007)
28
Impurity seeding confinement
Energy confinement (H98) decreases with density
(fGW) and radiation (frad)
JET
ITER
29
Energy confinement degradation
JET
Density (fuelling) scan
50
Normalised energy confinement (H98) reduced with
line average density as it approaches the density
limit (nGW)
H98 also reduced by 15 after a Type-I to
Type-III ELM transition
Radiation (impurity seeding) scan
H98 reduced with radiative fraction
Caused by reduction of pedestal temperature and
pressure
50
Since Wped 1/3 W, hence a 50 drop in Wped
means a 15-20 drop in H98
M.Beuskens et al, submitted to NF
30
Divertor heat loads

• Introduction
• JET ITER-like wall (ILW) ITER
• Divertor heat loads
• Divertor geometry and IR
• ELM filaments
• H/D/He comparison
• Impurity seeding
• Ne vs N2
• Effect on inter-ELM vs ELM heat loads
• Effect on H98 Zeff
• Disruptions
• TQ heat loads
• Conclusions Implications for ITER

31
Type-I ELM structure on divertor
Heat load imprints of individual ELM filaments
Imprints of individual filaments
Pre-ELM magnetic equilibrium
?t 0 µs ?t 85 µs ?t 170 µs ?t 255
µs ?t 340 µs ?t 425 µs
F
SOL
PFR
Pre-ELM separatrix
toroidal direction
target coordinate
Snapshot of IR camera displaying the divertor
target plates during ELM power load

• Near separatrix heat load profile roughly
  similar between and during ELMs
• Heat load imprints of single filaments resolved
  in the far scrape-off layer

Spatial resolution is 1.7mm, picture integration
time 20 us
Courtesy of T. Eich
32
Convective (small) ELM profile
ELM (40 kJ), DW/W 3, (1.0MA/1.1T), n 0.5,
conductive / convective 1
Averaged ELM
?q 8 mm (mapped to outer mid-plane)
Single ELM, smooth radial decay
x 1
MW/m2, log scale
max
pre
SOL
?q 8 mm
DW/W 3
target coordinate (m)
MW/m2
pre ELM
target coordinate (m)
PFR

• For convective ELMs, little or no broadening
  wrt. inter-ELM profile
• no ELM structure observed

time(s)
Courtesy of T. Eich
33
Conductive (large) ELM profile
ELM (400 kJ), DW/W 10, (2 MA / 2T), n 0.23,
conductive / convective 7.5
Averaged ELM
ELM filamentary structure becomes visible for
larger ELMs, at lower ?
?q 6 mm
x 1.5
MW/m2, log scale
Single ELM, ELM structure visible
?q 4 mm
max
pre
SOL
DW/W 10
pre ELM
target coordinate (m)
MW/m2
target coordinate (m)

• For conductive ELMs, broadening of near-SOL
  profiles by 50 wrt. inter-ELM profile
• Movement of peak heat flux position
• A lot more power to far-SOL hence limiter

PFR
time(s)
Courtesy of T. Eich
34
Type-I ELM profiles vs ELM size
DW/W 4

• This complexity of ELM heat load profiles
  generally increases with relative ELM size
• Comparison of two typical ELMs with relative ELM
  size of
• 4 (top)
• 9 (bottom)
• Note the rapid rise of far-SOL heat load, which
  explains the broadening of the ELM-integrated
  profile

DW/W 9
Courtesy of T. Eich S. Devaux
35
Imprint of ELM filaments

• Mapping of magnetic field lines on the divertor
  target

• Striations correspond to intersection of B field
  lines with the divertor target.
• Striations are interpreted as a result of plasma
  filaments ejected radially at the outer midplane.
• Hence, striations with different radial positions
  on the target correspond to filaments released at
  differtent toroidal locations at the outer
  midplane.
• Quasi Mode Number (QMN)

• B-field lines linking outer midplane to outer
  target

Courtesy of S. Devaux
Top view
36
Quasi mode number (QMN)

• Heat load striations appear
• in the early phase of the ELM crash,
• everywhere on the target, and
• only evolve during the rising phase
• Statistical analysis over large sample of ELMs
  (from a single pulse), focusing on the rise phase
  (up to max heat load)

• This phase is split into 2 stages
• 1st stage 0 0.4 ?IR
• ? QMN increases from 5 to 20
• 2nd stage 0.4 1 ?IR
• QMN saturates at 20
• energy delivered by each filament increases

Courtesy of S. Devaux
37
IR imprint on outer limiters
Difference between ELM and pre-ELM frames
ELM heat load superimposed on ambient background
66515
DWELM 200 kJt 7.6 sExp. time 300 msFrame
time 7.8 ms

• Impact of ELM filaments on outboard limiters
  observed with infra-red camera
• Several imprints observed on each limiter, none
  on the inner or upper plates

W. Fundamenski, M. Jakubowski, ITPA Garching May
2007, P. Andrew et al., EPS 2007
38
IR imprint field aligned
68193, 57 s
ELM filaments follow pre-ELM magnetic field lines
in the poloidal-toroidal plane
Also observed on the upper dump plates in high
triangularity plasma equilibrium
39
Fast-visible images of filaments

• Exposure time 33 ms
• Ten successive frames showing ELM-filaments
  striking the upper dump plate
• Less contact at outer limiter

Type-I ELM
40
IR imprint on upper dump plate
Post-ELM
Pre-ELM
Courtesy of G. Arnoux

• Wide angle IR image during an ELM
• Combined with EFIT reconstruction
• Helical stripes on upper dump plate
• Closely aligned with local magnetic field
  smaller pitch angle than at omp

41
ELM heat loads on outer limiter
DW/W 4
DW/W 9
DT/DW ? (DW/W)a, a 0.35 0.65

• Maximum temperature rise on outer limiter
  decreases with outer gap (distance between
  separatrix and outer limiter)
• When normalized by the size of the ELM, it
  increases roughly as the square root of the
  normalized ELM size
• Hence, larger ELMs deposit a larger fraction of
  their energy on the first wall, consistent with
  previous observations, and with divertor
  measurements

Courtesy of M. Jakubowski and T. Eich
42
Main Chamber Fluxes and ROG

• Determination of erosion sources, carbon
• influx and carbon concentration
• Impact of wall clearance on the primary C
  erosion
• Increase of C and Be erosion flux at midplane
  with
• reduction of wall clearance (spectroscopy)
• No variation in deposition on the QMBs
• Confirms step-wise transport

ROG 6 cm
ROG 4 cm
ROG 8 cm
C II
Be II
43
Simulations of filament motion
density, pressure vorticity
Radial distance
Radial distance
O.E.Garcia, N.H. Bian and W.Fundamenski., Phys.
Plasmas (2006)
44
Expression for ELM energy to wall on JET
Interchange driven amplitude scaling with
convective ion losses combined with
moderate-ELM (DW/W 5, DW/Wped12) e-folding
length, yields so that fraction of ELM energy
to wall can be approximated as where Dped is
the pedestal width and DSOL is the
separatrix-wall gap. eg. when DW/W reduced by a
third, then (Wwall/W0) 10 for 3 cm gap, see
below.
W.Fundamenski et al, PSI 2006 subm. to
J.Nucl.Mater
45
Divertor heat loads D, H, He

• Dedicated hydrogen and helium plasma
  experimental campaigns wrt D references
• Divertor heat load profiles measured in
  identical D, H and He plasmas
• Ion mass and charge have a modest effect on
  inter-ELM profiles

D
H
He
Courtesy of T. Eich
46
Divertor heat loads D, H, He

• Ion mass and charge have a pronounced effect on
  ELM ELM-average profiles

D
H
He
DW/W 5
DW/W 4
DW/W 4
D
H
He
Courtesy of S. Devaux
47
1-D kinetic (Vlasov) ELMs
Ion-electron Vlasov simulation
Power deposition on outer target
D
Field-free Vlasov simulation consistent with
analytic theory (free streaming ions)
Plasma potential
Electron momentum Ion momentum
Distance along field line
Courtesy of S. Devaux
48
1-D Kinetic (PiC) ELMs
JET 62221, WELM 0.4 MJ TELM 1.5 keV
Power loads to the divertor from the experiment
(62221, T.Eich), PIC simulation (D.Tskhakaya,
JNM 2009) and analytic fit using free streaming
ion result (Fundamenski, PPCF 06).
Courtesy of D. Tskhakaya
49
1-D multi-fluid transients

• ESELSOL1D
• plasma density

Courtesy of E. Havlickova
50
Impurity seeding

• Introduction
• JET ITER-like wall (ILW) ITER
• Divertor heat loads
• Divertor geometry and IR
• ELM filaments
• H/D/He comparison
• Impurity seeding
• Ne vs N2
• Effect on inter-ELM vs ELM heat loads
• Effect on H98 Zeff
• Disruptions
• TQ heat loads
• Conclusions Implications for ITER

51
Ne / N seeding of ELMy H-mode
Matrix scans executed in D fuelling Ne / N
seeding to explore potential for mitigating
target loads.
?
Large difference in N versus Ne input required
for similar impact on power exhaust (heat loads).
?
D fuelling tends to raise plasma density, while
seeding generally lowers it, ie reduces
confinement.
?
Courtesy of G. Maddison and the impurity-seeding
experiments team
52
Inter-ELM heat load strongly reduced
I IIIII IV
reference
seeded with strong heat load reduction
Divertor target heat loads measured by fast
infra-red thermography.
?
and further with Ne / N seeding.
Outboard inter-ELM peak heat load declines
substantially with D fuelling
?
Courtesy of impurity-seeding experiments team
Clear N shot-to-shot legacy effect.
?
53
Average heat loads strongly reduced
Ne
N
reference
seeded with strong inter-ELM heat load reduction
Inter-ELM and time-average peak heat loads on
outer target strongly reduced by Ne and N seeding
?
Outer divertor detached in-between ELMs in D
fuelled and Ne and N seeded plasmas above
?
Only a minor effect on peak ELM heat load with
Ne, and moderate effect with N seeding
?
Courtesy of impurity-seeding experiments team
54
Target power strongly reduced
reference
Ne
N
Inter-ELM power on outer target significantly
reduced with either Ne or N seeding.
?
Impact on ELMs is complex (appearance of
multiple/compound ELMs)
?
ELM size (and power on outer target) unaffected
by Ne, but roughly halved by N-seeding
?
Total (time-averaged) power on outer target
roughly halved by Ne and again halved by N-seeding
?
Courtesy of impurity-seeding experiments team
55
Radiation redistributed with Ne / N
Ne
N
reference
seeded with strong inter-ELM heat load reduction
In all three cases, the inter-ELM radiation is
maximum in the vicinity of the X-point
?
Gradual shift of the radiation zone from inboard
to outboard divertor legs in D only Ne N
seeded shots
?
Courtesy of impurity-seeding experiments team
56
Radiation not simply increased
total
total
divertor
divertor
reference
seeded with strong heat load reduction
Inter-ELM radiation fraction (total top, or
divertor bottom) only modestly increased by Ne /
N seeding.
?
At higher fuelling, radiation can actually fall
with stronger seeding owing to drop in plasma
density.
?
Courtesy of impurity-seeding experiments team
57
Performance modestly reduced
reference
seeded with strong heat load reduction
Large reduction of target heat loads comes at a
small cost in terms of confinement / purity ( 10
- 15).
?
Line-averaged purity tends to be compromised more
at lower fuelling ( lower plasma density).
?
Courtesy of impurity-seeding experiments team
58
Performance modestly reduced
reference
seeded with strong heat load reduction
Both seed species moderate target loads for
generally a small penalty in terms of confinement
/ purity.
?
Line-averaged purity tends to be compromised more
at lower fuelling ( lower plasma density).
?
Courtesy of impurity-seeding experiments team
59
Disruptions

• Introduction
• JET ITER-like wall (ILW) ITER
• Divertor heat loads
• Divertor geometry and IR
• ELM filaments
• H/D/He comparison
• Impurity seeding
• Ne vs N2
• Effect on inter-ELM vs ELM heat loads
• Effect on H98 Zeff
• Disruptions
• TQ heat loads
• Conclusions Implications for ITER

60
Heat load during thermal quench
Fast ( 1ms) infra-red measurements of heat loads
during TQ
Courtesy of G. Arnoux, A. Loarte, V. Riccardo
61
Heat load during thermal quench

• Clear observation of plasma (as opposed to
  radiation) load on limiters during TQ
• Comparable time scales for divertor and limiter
  peak heat loads ( 1 ms)
• Comparable energy fraction deposited on divertor
  and outer limiter (15)
• Spatial broadening and double peaked structure of
  outer limiter heat load profile
• Indicative of toroidally asymmetric plasma
  structure

Courtesy of G. Arnoux, A. Loarte, V. Riccardo
62
Conclusions

• Introduction
• JET ITER-like wall (ILW) ITER
• Divertor heat loads
• Divertor geometry and IR
• ELM filaments
• H/D/He comparison
• Impurity seeding
• Ne vs N2
• Effect on inter-ELM vs ELM heat loads
• Effect on H98 Zeff
• Disruptions
• TQ heat loads
• Conclusions Implications for ITER

63
Conclusions
64
THE END
65
Material migration
Courtesy of S. Brezinsek,
66
Be Migration after Fresh Be-Evaporation

• Study Be migration in a carbon-dominated tokamak

• Be evaporation to cover first wall and
• suppress C
• Fixed magnetic configuration (HT) with
• large wall clearance
• Shot-by-shot evolution of Be and C

Carbon concentration (L-mode)
initial value
plasma edge
plasma centre
K. Krieger et al. JNM 2009 / PFMC 2009
67
Evolution of Be Wall sources
mid-plane first wall (H)

• Erosion flux from calibrated Be II line
  intensity (527nm) assuming S/XB8
• Effect of 10 increased Be coverage is only
  30
• Depletion of Be layer ?t½ 400s
• ? No significant influence of gap wall
  separatrix
• Total Be erosion (140m2) 0.02-0.03g? local
  Be source ltlt avg source

wall gap ? L-mode high ? L-mode low ? H-mode low
68
Material Migration in Symmetric Magnetic
Configurations
JET MKII HD divertor
Diagnostics

• Post mortem analysis
• Tracer experiments
• Deposition monitors
• Visible spectroscopy

SOL
SOL
PFR
Closest divertor configurations with respect to
ITER!
Study of the complex divertor migration in a
series of consecutive discharges
69
Typical observation at the inner QMB
Material Migration in the Divertor
Deposition on the QMB in the inner divertor
pump duct entrance
lt
S. Brezinsek et al, EPS 2008

• Magnetic configuration
• Power and heat load at the strike-points steady
  state and transients
• Particle fluxes at the strike-points ions and
  neutrals
• Local surface conditions soft hard carbon
  layers and CFC

70
Magnetic Configuration
QMB in the private flux region located below tile
5
QMB located in the inner duct entrance

• No deposition after extraction
• of this QMB
• after the last opening
• Surface temperature increased
• due to operation with OSP on tile 5

A. Kreter et al. JNM 2009
H.G. Esser et al. JNM 2009

• Carbon deposition on inner QMB reflects
  re-erosion at strike-point location
• Line-of-sight transport in the inner divertor

• Portion of carbon re-eroded on tile 3 and
  transported towards the PFR
• Erosion of deposited carbon in the closed PFR
  due to deuterium neutrals

71
Typical observation at the inner QMB
Material Migration in the Divertor
ELM induced erosion Non-linear effect
Other parameters constant

• Magnetic configuration
• Power and heat load at the strike-points steady
  state and transients
• Particle fluxes at the strike-points ions and
  neutrals
• Local surface conditions soft hard carbon
  layers and CFC

72
ELM-Induced Enhanced Erosion
Inner louvre Deposition per ELM vs. ELM energy
Processed exp. data
16
10
15
10
10x less
C deposition per ELM atoms/cm2
Arrhenius-type equation

• Pre-conditioned
• target (tile 4)
• ELM-resolved
• deposition on QMB

14
10
Physical sputtering
13
10
0
100
200
300
400
A. Kreter et al. JNM 2009, PRL2009
ELM energy ?WELM kJ

• Thermal decomposition of hard carbon layers
  under ELM impact (lt500kJ)
• This enhanced re-erosion due to ELMs is observed
  for (carbon) layers
• Leads to redistribution of deposited material in
  the divertor leg

73
Ablation of Carbon Layers by Giant ELMs
Spectroscopy
Bolometry
800
69817,18 70225,6,8
700
600
500
0.51
?Erad ELM (kJ)
400
300
200
100
0
220
320
420
520
620
720
820
920
R. Pitts et al. JNM2009 A. Huber EPS 2008
?W ELM (kJ)
ablation
thermal decomposition

• VT-like configuration with conditioned vertical
  targets
• Repetitive unfuelled H-mode discharges with ELMs
  up to DWELM1.2 MJ

• Giant ELMs above DWELM0.5 MJ lead to
  destruction of hard surface layers
• Cross-divertor transport of ablated particle
  clusters (including Be)

74
Fuel retention
Courtesy of T. Loarer,
75
Gas balance on JET
Regen. of the Divertor cryopumps before and after
the session ? Total gas pumped (acc.1)
Repeat sets of identical pulses (no intershot
conditioning) to avoid/limit the possible
contribution of the history effects
Plasma
Injection Pumped Short Term Ret Long Term
Ret
Regen. from Divertor cryo Pumped during plasma
outgased in between pulses
76
Gas Balance Studies at JET
Regenerate cryogenic pumps before and after expt.
? collect total pumped gas
Repeat sets of identical discharges
Plasma
Calibrated Particle Source (Gas, NBI, pellets)
Wall Retention Long Short Term
Divertor cryo-pumps
Injection Pumped Short Term Ret Long Term
Ret
Total recovered from cryo-regeneration Pumped
intershot outgassing over 800s (assumed equal to
Short Term Ret )
T. Loarer PFMC 2009
Dedicated series of experiments L and H mode in
2007-2009 as references for ILW
77
Type I ELMy H-mode Experiment
HT3 configuration
PTOT (MW) NBIICRH13MW
DWELM 100 kJ 60 Hz
Da (in)
Da (out)
Time (s)

• Short term retention limited to fast
  reservoir
• Recovered in between pulses (outgasing)
• Long term retention Co-deposition and
  implantation
• Slow process compared to short term over 5-10
  sec

78
Experiments in HT3 Configuration

• Long term retention in carbon-dominated JET
  depends on plasma scenario
• Increase of retention from L-mode to moderate
  type I ELMy H-mode by 33
• Moderate impact of the limiter phase in the
  fuel retention experiments

• Main mechanism co-deposition of fuel by carbon
  due to material migration
• increase of the ion flux to the wall
• increase of carbon erosion in the main chamber
• carbon flux increased

79
Recycling (Da) and retention
Type I ELMy H-mode Ip/BT2.5MA/2.7T
L mode and Type III H-mode
Type I ELMy H-mode Ip/BT2.0MA/2.4T
- No clear correlation of the retention with CIII
extra carbon resulting from ELMs does not
appear and/or affect the retention
80
Extrapolating Da and retention
Increasing ion flux at the inner divertor leads
to an enhanced retention
81
Conclusions

• Gas balance integrated over a session
• - Allows for a good accuracy (1)
• Avoid possible contributions from
  history/conditioning effects
• Reconstruction of particle fluxes during plasma
• Extrapolation to high power discharge (same
  plasma shape).

• - The higher recycling flux the higher retention
  results for attached plasmas
• - Within a factor of 2 the recovery is constant
  in the range 1-3x1022D
• ?No major contribution on the overall long term
  retention
• - Reference dischagres in carbon before the ILW
  (Be/W) and quantify the benefit (?) of a full
  metal wall compared to C.

82
Short term retention
Recovery after the pulses

• Small fraction recovered but gt plasma content
  0.5x1022D
• (70m3, ltnegt5-6 1019m-3)

• Independent of inventory accumulated during the
  pulse and previous pulses

• Except for disruptions, this amount is constant
  and independent of Ip, BT, ne, Pin, Ginj, Wdia
  (plasma scenario)

Within a factor of 2 the recovery is constant in
the range 1-3x1022D No major contribution on the
overall long term retention
83
General results L and H-mode
High triangularity shape (HT3)
84
General results L and H-mode
High triangularity shape (HT3_L)
HT3_R (2.5MA)
85
Edge-SOL Modelling

• Introduction
• JET-ILW-ITER, goals challenges of ILW
• Divertor heat loads
• Divertor geometry and IR, ELM filaments, H/D/He
  comparison
• Impurity seeding
• Ne vs N2, effect on inter-ELM vs ELM, H98 Zeff
• Disruptions
• Description of DMV MGI, TQ heat loads, CQ
  runaways
• Material migration
• Be evaporation, QMBs, C-spectroscopy detachment
• Fuel retention
• Outline of gas balance experiments
• Modelling
• EDGE2D/EIRENE, ESELSOL1D, kinetics
• Conclusions

86
2-D multi-fluid DCN
NITROGEN RADIATION (Wm-3)

• EDGE2D-EIRENE simulations of (D,C,N) JET plasmas
  successfully converged
• Essential for understanding the interplay
  between C and N radiation in Nitrogen seeded
  plasmas
• Initial results suggest that N radiates locally
  near the puff location and selectively enhances
  detachment at inner or outer targets
• Reproduce experimental trends in increase of
  Z_eff with seeding at low density, and little
  effect at high density
• Similar simulations with Ne underway

N puff
No puff
OT puff
IT puff
Power to OT
Power to IT
Power to wall
N radiation
Other volumetric losses
Heat load at IT
Heat load at OT
Courtesy of D. Moulton
87
2-D multi-fluid u vs. t profiles
Scaling of upstream SOL widths with target
widths Relation between heat load width _at_
target and electron temperature _at_ upstream,
generally larger than the classical ratio of 2/7 !
Courtesy of D. Moulton
88
1-D multi-fluid transients

• ESELSOL1D
• temporal profiles
• cross-field sources
  mid-plane values target values
  parallel losses

Courtesy of E. Havlickova
89
1-D multi-fluid transients

• ESELSOL1D
• parallel profiles

Courtesy of E. Havlickova
90
Edge-SOL turbulence not anomalous
Edge/SOL turbulence is no longer anomalous.
Predictive capability in sight.
Recall that anomalous abnormal, irregular, not
understood
Ironically, it is the absence, rather then
presence of turbulence which now appears
anomalous.
Theres just one more thing that bothers me
How do I get this H-mode ??!
We know who did it. We still dont know how.
flow shear poloidal toroidal rotation
magnetic shear, X-point geometry, ion orbit
losses, bootstrap current,
The Holy Grail of tokamak theory ! and main
obstacle in predicting tokamak plasma exhaust
(ITER)