Alberto Loarte European Fusion Development Agreement Close Support Unit – Garching

Alberto Loarte 9th ITPA Divertor and SOL Physics Meeting IPP-Garching 7-10/5/2007 1

Report on ITER Design Review Sub-group on : Heat and Particle Loads to

in-vessel components associated with limiter and X-point operation, TF

ripple, H&CD systems, ELMs, disruptions, VDEs, Marfes and runaway

electrons in ITER

Alberto LoarteEuropean Fusion Development Agreement

Close Support Unit – Garching

Acknowledgements: A. Grosman, P. Stangeby, G. Saibene, R. Sartori, M. Sugihara,

W. Fundamenski, T. Eich, P. Snyder, V. Riccardo, G. Counsell, R. Pitts, B. Lipschultz,

P. Andrew, G. Pautasso, A. Leonard, G. Strohmayer, G. Federici, A. Kirk, J. Paley,

M. Lehnen, B. Alper, C. Ingesson, etc.


ITER PID only specifies Pmaxlimiters = 15 MW (max 9 MW on one limiter)

Fluxes to limiter during Ramp-up/down (I)

Reference ITER ramp-up(/down) has long limiter phases up to Ip = 7 MA (10 MA) in which

plasma is limited by two limiters 180o apart (power loads & erosion)

2 limiter configuration and qlim = 5 lead to long

connection lengths in SOL (>> 200 m)

Magnetic shear + perpendicular transport simple “single exponential” power decay length (Kobayashi, NF 2007)Main Uncertainties

PSOL for all ITER reference scenarios (ramp up/down with heating)

Scaling of SOL transport with Ip and R (JET extrapolation for Kobayashi NF 2007)


ITER power and power fluxes estimated with B2-Eirene for a range of burn conditions (mainly QDT = 10) to maintain detachment but weak physics basis

for SOL transport and main chamber fluxes

Fluxes to main wall and divertor during diverted operation (I)

near SOL transport p = 5 mm (close to most pessimistic scalings 4 mm) for Ip = 15

MA (6 mm for scenario 3 (hybrid) and 8 mm for scenario 4 (ITB) if p ~ Ip-1)

qII = 570 – 760 MWm-2 for PSOL = 100 MW (typical for scenarios 1 & 2)

H-mode scenarios 1 & 2 wall = 5–20 cm qIIwall < 0.04 MWm-2 lIB

L-mode scenarios 1& 2 (PSOL=35 MW, L=2 H) & wall=5–20 cm qIIwall < 1.0 MWm-2 IIB

0.5o < < 15o FW load < 0.5 MWm-2 fulfilled but loads on edges ? (2mm steps) and edges of ports ?

Main uncertainties : In/out divertor power asymmetry (ballooning transport) Far SOL transport (wall fluxes) Scaling of p


Particle flux to ITER main wall expected to be > 1023 s-1 (> 1% of div)

Fluxes to main wall and divertor during diverted operation (II)

Lipschultz

0.00 0.05 0.10 0.15

105

106

L = 60 m, TSOL

= 10-20 eV, VSOL

= 30 - 100 m

qup

II qlow

II

qup

perp qlow

perp

Distance to Separatrix (m)

q II (W

/m2 )

104

105

qperp (W

/m2)

0.00 0.02 0.04 0.06 0.08 0.10 0.12 0.140.0

2.5

5.0

7.5

10.0

12.5

15.0

17.5

20.0

Distance to active separatrix at midplane (m)

An

gle

of

inci

de

nce

(o )“Scarce” data & ITER B2-Eirene modelling : n (sep ~ 5 cm) = 0.4 - 1.0 1019 m-3

vSOL = 30 – 100 ms-1

TSOL = 10 – 20 eV SOL = 4.5 – 21 cm

qII < few MW II B on (outer) first wall local particle & power fluxes on edges and edges of ports ?

qperp < 0.3 MW m-2 < OK for FW panels


w/o convection

with convection

P//=94kW (/20MW)P//=88kW

Q//=0.4|VRF|necs

•Integrated losses : typically 100kW for 20MW injected (low density case)

•Localised peak at 10MW/m2 ; average

~2.5MW/m2

•These results depend crucially on the density value in the first cm in front of the wall (far-SOL transport ?) • Sheath rectification may reach 2-3 kV

Sputtering of surfaces!

Fluxes associated with heating systems and steady-state non-toroidal asymmetries

NBI shine-through limit ne > 3.7 1019 m-3 (30% nGW) for 0.5 MWm-2 (edges ?) but no

local ionisation or first orbit losses included Ripple losses expected to to lead to less than 0.3 MWm-2 (3-D effects ?) Effect of ELM RMP coils on power deposition assymetries ? ICRH (and LH) can lead to large power fluxes on PFCs near and far field (not included

in PID) EFDA Task TW6-TPHI-ICFS2 (L. Colas, CEA)


In transient events 1 cm SOL field line touches first wall for 1 s

Plasma Position and Shape Control

If plasma in H-mode then depending on location of contact plasma stays in H-mode or H L transition

H-mode H L350 MJ 175 MJ

qIIcontact-wall 77 – 102 MWm-2

pcontact-wall 0.5 cm

tcontact-wall 1 s

qIIL-H contact-wall > 105 - 140 MWm-2

pL-H 1.0 cm

pL-H < 1.7 s

sin = 1-8 10-2

sin = 1.5-2.5 10-1

in “minor disruptions” separatrix can touch dome for ~ 1 sqII in,dome ~ 380 - 570 MWm-2


PID estimates of ELM loads for ITER carried out on simplified experimental basis

Fluxes to main wall and divertor during ELMs (I)

Specified loads are of the right magnitude but can be improved to include ELM physics understanding (time dependence, in/out asymmetries, relation “” vs WELM/Wped

Sugihara, ITER_D_22JYYU, 2005

700-950 MWm-2 3.5-4.7 GWm-2


Fluxes to divertor during ELMs (II)

222

exp1)(ttt

tqELM

WELM < 30 MJ

Adiv,ELM = 4 m-2

Broadening < 1.5

rise,ELM = 200-500 s

down,ELM = 1-2 rise,ELM

Loarte, PPCF’03

Eich, PIPB’07

Eich application of

Fundamenski PPCF’06

Eich, PIPB’07


Fluxes to divertor during ELMs (III)

TPFdiv,ELM < 1.5

Ein,ELM/Eout,ELM = 1-2

Divertor ELM load near separatrix ~ toroidally symmetric but strong in/out asymmetries

Eich, PRL’4

Loarte, PPCF’03 from Leonard JNM’97

DIII-D

Eich, JNM’07

Eich, JNM’07


Fluxes to divertor during ELMs (III)

Wdiv,ELM 5 MJ 30 MJ

Einmax (MJ) < 3.3 < 20

Eoutmax (MJ) < 2.5 < 15

Einmin (MJ) < 2.5 < 15

Eoutmin (MJ) < 1.7 < 10

qinmax (GWm-2) < 6.2 < 37.5

qinmin (GWm-2) < 1.9 < 11.3

qoutmax (GWm-2) < 4.7 < 28.1

qoutmin (GWm-2) < 1.3 < 7.6

risemax (s) 500 500

risemin (s) 200 200


Part of WELM is reaches the main wall PFCs formation and ejection of filaments

Fluxes to main wall during ELMs (I)

0.05 0.10 0.15 0.200.0

0.2

0.4

0.6

0.8

1.0

1.2DOC-L = 0.27 1.2MA q

95 = 3.1

2MA q95

= 3.7

2MA q95

= 4.6

3MA q95

= 3.1

WE

LM

IR/

WE

LM

WELM

/Wped

Model of qII(t) for detached filaments developed by Fundamenski (PPCF’06) and validated with JET data (Pitts NF’06) Application to ITER

JET-Eich, PIPB’07

MAST- Kirk, EPS’06

AUG- Herrmann –PPCF’06

?


qII in filament estimated with model by Fundamenski required input to model : nfil, Tfil, <vELM/cs,ped> & distance from sep @ filament detachment

Fluxes to main wall during ELMs (II)

ELM fluxes to main wall (beyond second sep) only on outer wall Power reaches the wall in filamentary structures (for ITER Snyder results in NF’04) :

distance between filaments (m) = 15/n (if no break-up and all become unstable) filament poloidal width (m) = 3/n (rough estimate) Decay length of filaments in “limiter” shadow ~ Llim/LSOL ?

Outstanding issues : Relation VELM & WELM vELM/cs,ped = 10-2 WELM/Wped or 1.5 10-3 (WELM/Wped)0.5

nfil, Tfil Rfil at instant of detachment & ELMWall (~ ELM

div ?) wall

ELM sputtering

0.0000 0.0005 0.0010 0.0015 0.0020

0.1

1

10

From modelling by W. Fundamenski

5 cm from sep 10 cm from sep 15 cm from sep

VELM

/CS

q II (G

Wm

-2)

nfil = 7.5 1019 m-3, T

fil = 5 keV, detachment from R

ped = - 5cm

0

200

400

600

800

1000

qII upper-B

e-walllow

er-Be-w

all (MW

m-2)

Herrmann-AUG


PID specifications generally in line with current evidence from disruption loads but need to be refined to incorporate latest findings on divertor/wall

loads

Fluxes to main wall and divertor during Disruption thermal quench (I)

Classification of loads per disruption type (ideal MHD limits, etc.) and scenario Disruptions in limiter phase are absent in specifications Toroidal and in/out asymmetries ? Radiation during thermal quench ?


Fluxes to main wall and divertor during Disruption thermal quench (II)

Plasma energy at t.q. typically less than 40% expected from H98 = 1 (Size scaling ?)

Dedicated experiments at JET in 2006/2007 show that Wt.q. < 0.4 W (Type I H-mode) for density

limits, radiative limits and NTM driven disruption

JET-Riccardo NF’05

MAST-Counsell

t.q. timescale has large scatter associated with MHD activity but similar to ELMs large amount of energy

reaches PFCs after qmax


Fluxes to main wall and divertor during Disruption thermal quench (IV)

Broadening of power width causes energy deposition IIB everywhere on PFCs (TPF < 2) significant amount of energy deposited outside divertor

Disruptions in divertor conditions triggered by ideal MHD limits and in limiter seem completely different many aspects

(P. Andrew, PSI’06, Riccardo NF’05, Finken NF’92, Janos JNM’92, TFR JNM’82)

JET-Paley-PhD Thesis ‘07


Fluxes to main wall and divertor during Disruption thermal quench (V)

Ideal MHD limit disruptions can lead to large interactions with inner-wall or outer wall not seen in other disruption types Not included in PID for scenario

4 (ITB) implications for ITER ?

PNBI (X 10 MW) PICRH (X 10 MW)

Prad (X 100 MW)

JET Pulse No. 69816 ITB grad-P disruption

Wdia (MJ)

Ip (MA)

Mode-lock (a.u.)

n = 2 (a.u.)

n = 1 (a.u.)


Disruptions in limiter phase

Not considered in PID but potentially serious because of lack of broadening of power footprint for limiter disruptions (t.q. < 1.5 s.s.)

EOL-ramp-up : Ip = 7 MA, if Pinp = 5 MW Wplasma (ITER-89) = 15 MJ

BOL-ramp-down : Ip = 10 MA, if Pinp = 7 MW Wplasma (ITER-89) = 24 MJ

TEXTOR-Finken NF 1992

For p > 2 cm Aeff,limiters = 2.5 m2 (H. Pacher)

Disruptions during limiter phases may cause loads > 6 – 10 MJm-2 with t.q. ~ 1 msMajor issue for power fluxes during VDEs needs to be confirmed

Janos-TFTR-JNM 1992

Normal discharge

Disruptive-normal discharge


Large discrepancy between PID specifications and new proposed specifications in Sugihara NF ‘07

Energy Fluxes to main wall and divertor PFCs during VDEs (I)

~ 134 MJm-2s-1/2

Possible Realistic scenario Plasma drifts towards wall in H-mode At some point L-mode transition (H-modes

with X-point behind target at JET) Wplasma (30-50 % of Wplasma) deposited on wall in

t < L-mode with p ~ 1 cm Plasma is in contact with wall in limiter L-mode

PSOL = 100 MW + dW/dt Plasma disrupts in limiters configuration when

q ~ 1.5-2 with Wplasma > 100 MJ


PID specifications for Marfe loads in ITER (physics model ?)

Energy Fluxes to main wall and divertor PFCs during Marfes (I)

Three types of “Marfes” expected in ITER (L-mode Plasma) :

Inner-wall Marfe Potentially steady-state Prad < PSOL (100 MW) (<0.15MWm-2>)

X-point Marfe Potentially steady-state Prad < PSOL (100 MW) (<0.15MWm-2>)

Pre-thermal quench divertor Marfe short-lived (< 0.1 s in JET and AUG) period in

which ~ 0.25 of Wplasma (H98 = 1) can be radiated Prad ~ 900 MW (<1.3 MWm-2>)

Pre-thermal quench limiter Marfe short-lived (< 0.1 s in JET) period in which a

fraction (?) of Wplasma (H89 =1) can be radiated Prad < 240 MW (<0.35 MWm-2>)

For all cases radiation peaks near X-point region or inner-wall

Main issue is to determine realistic timescales and peaking factors of radiation on wall

due to Marfe for ITER


Examples of Marfes at JET

Energy Fluxes to main wall and divertor PFCs during Marfes (II)

Steady-state limiter MarfeA. Loarte Memo to RI-mode working group’99

Transient limiter MarfeJ. Wesson-Science of JET’99

~ Steady-state X-point MarfeA. Huber-PSI’06


Conclusions

Many of the PID specifications for PFC loads in ITER are not far

from expectations from latest experimental/model results Other are in disagreement with present evidence and/or absent A detailed review and update of PID specifications is needed &

will be carried out as part of the design review Contributions from ITPA groups and collaboration with ITER-IT

will be essential to do this review

Expected loads in ITER will determine fine details of PFC

construction (edge shadowing, etc. ), overall PFC configuration &

will have implications for the use of diiferent PFM in various areas

of the device


Other transients following plasma disturbances and noise in feedback/measurement system

Plasma Position and Shape Control (II)


Even if position control is recovered in 10s there are ~ 1s phases in which separatrix gets dangerously close to areas designed for low power loads

Plasma Position and Shape Control (I)

SOB Minor disruptionAppendix E

qII in,dome ~ 380 - 570 MWm-2 for ~ 1 s

Control of plasma in ITER can lead to fluxes IIB on PFCs > 100 MWm-2 for timescales ~ 1s and possibly fast H-L transitions


Most tokamaks find that close to 100 % of magnetic energy at t.q. is radiated with the exception of Alcator C-mod (< 25%)

Energy Fluxes to main wall and divertor PFCs during current quench

Wmag = ½ Lp Ip2

After t.q. p= 0 (0.2-pre), li~ 0.5 (0.85-pre), Ip <18 MA (15 MA-pre) Wmag < 1.85 GJ

a

Most of Wmag V.V. and in-vessel conducting structures Wc.q.PFCs < 315 MJ

For fastest timescale of c.q. in ITER ~ 16 ms (exponential) 36 ms (linear)

qradmax < 80 MWm-2

critical assumptions : Wc.q.PFCs < 315 MJ, 100% radiation & peaking factor < 1.4


Mitigation of disruptions by massive gas injection is a promising scheme but may lead to large fluxes on PFCs specification of

wall loads for ITER not yet in PID

Energy Fluxes to main wall and divertor PFCs during mitigated disruptions

Sugihara NF’07 assumes qrad = 0.5 – 1.0 GWm-2 in 1 ms

Estimates from Whyte for ITER predict <qrad> ~ 4 GW/m2 for trad < 200 s

Whyte PRL’02


Runaway generation mechanisms for ITER like disruptions conditions studied in detail but runaway losses and dynamics

are worse known specification of local loads for ITER ?

Runaway electron fluxes on PFCs (I)


Runaway electron fluxes on PFCs (II)

Runaway electrons are not generated if q95 < 2-2.5 before current

quench (no r.e. in full VDEs but likely in disruptions) Runaway electrons are lost to PFCs by MHD turbulence when qedge =

3, 2 touches the wall Runaway beam has a peaked current profile ar.e. ~ 0.25 aplasma and is

vertically unstable Runaway impact point determined by vertical instability of column and

narrow e-folding length (~ few mm) Aeff ~ 0.5 m2 (if toroidally

symmetric) Runaways are lost in bursts of ~ 100 s over ~ 5-10 ms timescales

(JET and JT-60U)

Unclear whether all these facts are taken into account in present PID

specifications or not

Documents

Alberto Loarte European Fusion Development Agreement Close Support Unit – Garching