Tokamak edge physics and plasma-surface interactions
Richard A. Pitts
Centre de Recherches en Physique des PlasmasEcole Polytechnique Fédérale de Lausanne, Switzerland
Association EURATOM-Swiss Confederation
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts2 of 34
“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
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts3 of 34
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.
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts4 of 34
Divertor and SOL physics
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts5 of 34
Terminology: limiters and divertors
Core plasma
Core plasma
Scrape-off layer (SOL) plasma: region of open field lines
Divertor targets
Limiter
Vessel walls
Private plasma
Separatrix
LCFS
X-point
“Upstream”
Outermidplane
CL
OuterInner
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts6 of 34
JET #62218, H-mode, Ip = 3.0 MA, B = 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
e.g. Limiter and divertor phases in many JET shots
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts7 of 34
Basics – SOL width, n [1]• Any solid surface inserted into
a plasma constitutes a very strong particle sink
• In the high tokamak B-field:
<< ||
• Thin Debye sheath (D few 10’s m thick ) forms at the surface controls flow of particles and energy || B
e.g. L ~ 30 m (typical of JET): TLCFS ~ 100 eV, cs ~ 105 ms-1, D~ 1 m2s-1 (near SOL)
n~ 1.7 cm!!cf. minor radus = 2.0 m for ITEREven worse for energy – see next ……
• Quick and dirty estimate of n with diffusive approx. for cross-field particle transport (all ionisation inside LCFS): nv = -Ddn/dr ~ Dn/ n
v D/n , n= v
v|| cs ~ (kT/mi)1/2
Then, if = ||,
Dn /2
scL /||
2/1/ sn cLD
q, T, n
WALL
Limiter or
divertor plate
SOL
Main plasmaB
2L“Connection length”
“upstream”
Adapted from [1]
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts8 of 34
The problem with q• SOL width for power, q, is also small and is an important parameter of the edge plasma• As for particles, q 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 q 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 q = 5 mm, JET scaling gives q = 3.7 mm (cf. a=2.0 m)
− Very recent multi-machine scaling [8] gives q/R ~ constant
• Note also that the parallel power flux, q|| PSOL/q ~ as much as 1 GWm-2 in ITER
9/5 SOLq P15.04.0
959.05.0
uSOLq nqBP
Stored energy scales strongly with tokamak major radius, W ~ R4 [9]But power deposition area in the divertor Rq 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 today’s 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 today’s expts., where it is “useful”)
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts9 of 34
Power handling – ITER case (approx)
• 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
SOL
CORE PLASMA
(adapted from [10])
~100 MW
22.0~)/(2~ mBBRArea uq q||,u ~ 500 MWm-2
Magnetic flux expansion ~(B/B)u/(B/B)t ~4 for ITER outer divertor low field line angles at strike points (~3º)
+Target tilting in poloidal plane ( ~ 25º for ITER outer target)
q = 5 mm
20.2~)/)(sin(/)/(2~ mBBBBRArea tuq per target
q ~ 25 MWm-2 per target if no radiative (or other) dissipation
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts10 of 34
* rises as nu rises, finite electron heat
conductivity: (note: 0,e » 0,i)
allows parallel T gradients to develop Tt
decreases, but pressure balance maintained (p|| ~ 0) so that nt rises strongly ( )
ion ( 1/nt) decreases so that target recycling
increases strongly flux amplificationAs Tt , radiation loss increases Tt further
2/50||||||||, ,/ TKdsdTKq cond
2ut n
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: = sheath heat transmission coefficientpot = potential energy per incident ion
iieieeieuuucoll TTTnT ~~,~~,/2
collL /*
potstttsttt cnTcnq ||,
Target
T
n
Lu t
Low n, high T (high PSOL)
“Sheath limited”
ut TT
2/ut nn
Target
T
n
Lu t
Moderate n, T“High recycling”
3ut nn
2/1 ut nT Region of strong radiation losses
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts11 of 34
The route to detachment (2)At sufficiently low Tt, (< 5 eV), neutral ionisation rate < 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
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).
Target
T
n
Lu t
High n “Detached”
Recycle region
C-Mod, B. Labombard, et al., [11]
Adapted from [2]
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts12 of 34
Full detachment is a problem
JET, A. Huber, et al.
[12]
• 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
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
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts13 of 34
Divertor closure
• 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
JET, R. D. Monk, et al. [13]
Increasing closure
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts14 of 34
Target orientation• Parallel heat fluxes
significantly reduced for vertical cf. horizontal targets
• Underlying effect is preferential reflection of recycled deuterium neutrals towards the separatrix
Hotter plasma near separatrix
Increased ionisation near sep.
Higher nt, lower Tt
Higher CX losses
AUG, A. Kallenbach, et al. [14]
Cooler, less dense plasma
Separatrix
Pressure loss q||
Separatrix
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts15 of 34
Impurity seedingDIII-D, C. J. Lasnier, et al. [15]
JET, G. F. Matthews et al. [16]
D2 puff92 torrls-1 for 1.8 s
Ne puff12 torrls-1 for 0.1 s
Unfuelled
Strong D2 puff
Strong D2 +N2 puff
Strong impurity seeding also reduces ELM size but high price can be paid in confinement
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts16 of 34
ITER divertor achieves partial detachment
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
Inner strike pt.
Kirschner et al. [17]
Po
we
r lo
ad
(W
m-2)
Outer strike pt.
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts17 of 34
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).
To cryopumps
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 p, 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.
105/*, EHep
plasma
pump
eplasmaHe
pumpD
pumpHe
He C
C
nn
nn
/
2/ 2
*,Hep e.g. ITER: He prod. rate ~21020s-1
Max. divertor pumping speed ~200 Pa m3s-1 ~ 11023 He atom s-1
Cpump ~ 210-3 = 0.2%Typical acceptable He conc. in the core: ~4% He = 0.2/4 = 0.05 is minimum required. The values of and required for ITER have been achieved experimentally
*,HepHe
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts18 of 34
Plasma-surface interactions
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts19 of 34
ParameterJET MkIIGB(1999-2001) ITER
Integral time in diverted phase 14 hours 0.1 hours Number of pulses 5748 1 Energy Input 220 GJ 60 GJ Average power 4.5 MW 150 MW Divertor ion fluence 1.8x1027 *6x1027
1 ITER pulse ~ 6 JET years divertor fluence
1 ITER pulse ~ 0.5 JET years energy input
*Code calculation
Upscale to ITER is a big step Extracted from G. F. Matthews et al. [19]
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts20 of 34
ITER materials choices
• Be for the first wall−Low T-retention−Low Z−Good oxygen getter
Driven by the need for operational flexibility
• 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
• For D and DT phases:−Be wall, all-W divertor
To avoid problem of T-retention
W
CFCWhat are the issues associated with plasma-surface interactions?
Beryllium
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts21 of 34
Critical issues
Steady state erosion
Long term tritium retention
Material lifetime
Transient erosion(ELMs, disruptions)
Material mixing
Short and long range material migration
Redeposition
All strongly interlinked
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts22 of 34
Impurity migration
TransportErosion DepositionRe-erosion
=Migration
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts23 of 34
Chemical (carbon)
• 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
ITER divertor flux
D impact
Erosion: Physical and chemical sputteringPhysical
Adapted from Eckstein et al. [20] Roth et al. [21]
Current steady state divertor target erosion rates (ERO modelling) due to Yphys and Ychem estimated at ~0.4 - 2 nms-1 for ITER [17]
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts24 of 34
Erosion: transients, e.g. ELMs on the divertor
Important factor is max. Tsurf due to
arrival of short heat pulse (duration, t):
f = fraction of Wth lost during transient
Adiv, divertor wetted area
,k,C=density,conductivity,heat capacity ELM energy losses must stay below melting/sublimation/evaporation limit to avoid fast erosion (e.g. melt later loss)
2/1
max 2
Ck
t
A
fWT
div
thsurf
Time (s)
Important to measure Tmax
Federici et al. [22]
If transient ~ few m and target thickness ~ cm lifetime ~104 events~103 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
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts25 of 34
Erosion: ELM size must be mitigated in ITERMaterials 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 s, triangular pulse lengths (250 s up and down). How does this convert to an ITER tolerable ELM size [25]?
WELM = qELM A,in (1 + Eout/Ein) = 0.5 MJ/m2 1.4 m2 1.5 ~ 1 MJThis 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])
What do we know about ELMs at the divertor target?1) Rise time to peak Tsurf ~ || [26] (|| = cs/L ~ 250 s for ITER)2) Toroidal peaking factor ~ 1 [27]3) q,ELM ~ q,inter-ELM ~ 5 mm (ITER)4) A,IN = 1.4 m2, A,OUT = 1.9 m2
5) No ELM energy to main walls6) Strong in-out asymmetry in ELM power loading (not understood yet)
Eich, Pitts [28, 29]
JET
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts26 of 34
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
Transport creates and moves impurities
EDGE2D/NIMBUS
Bypass leaks
Escape via divertor plasma
Ionisation
D0 from wall ion flux or gas puff
CX event
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts27 of 34
20g Be (BeII)
450g C (CIII)
~250 kg/year if JET operated full time!Carbon migrates to remote locations forming D-rich soft layers (high T-retention)
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
~400g C22g Be
Likonen et al. [34]
Main chamber: source of
net erosion
• 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)
Strachan et al. [33]
Pitts et al. [4]Matthews et al. [32]
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts28 of 34
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 > 1
depending on the type of re-deposited layer
• Retention very hard to characterise in today’s 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.
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts29 of 34
Roth et al. [37]
C targets, Tsurf = 800ºC, chemical+physical sputtering
~0.02g/discharge
~2g/discharge
Tritium retention (2)• A 400 s QDT = 10 ITER discharge
will require ~50 g of T fuelling(cf. 0.01-0.2 g in today’s 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
• 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
ITER target [18] is a retention level of ~0.05 g/discharge ~7000 shots before major shutdown for T-removal
Accurate measurement of T-retention and the development of efficient T-removal methods will be critical for the success of ITER
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts30 of 34
Dust
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.
• Dust is seen in all tokamaks, especially with C walls• Not generally a concern in today’s 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
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]
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts31 of 34
Mixed MaterialsNo 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]:
Material property changes due to mixing
Effect on H-isotope retention
Effect on material erosion
• 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
• 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
• 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.
Preparations underway at JET to test a Be/W & Be/W/C wall mix from ~2011 [40]
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts32 of 34
References (1)
[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 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:
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
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts33 of 34
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
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts34 of 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
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts35 of 34
Reserve slides
Hungarian Plasma Physics and Fusion Technology Workshop, Győr, Hungary, 26/03/2008 R. A. Pitts36 of 34
Edge Diagnostics on ITER will be critical
Target plate heat and particle fluxes, Te, surface temperature, erosion rate
Neutral gas pressure
Cryopump inlet composition (H/D/T/He, CxHy)
Te ne Prad
Dust accumulation?
ne, Te, Prad, position of ionisation front, nHe, nD/nT, impurity and D, T influxes
Wall temperature and visible imageMain chamber gas pressure and gas compositionSOL neutral density (D/T)Impurity influxesSOL ne, Te profiles (but challenging)
See talk by A. H. Donne
Courtesy of A
. Kuku
shkin