Physics basis and design of the ITER full-tungsten divertor1 @2013, ITER Organization 55th Annual Meeting of APS Division of Plasma Physics, Denver, USA, 11-15 November 2013 IDM UID:

1 @2013, ITER Organization

55th Annual Meeting of APS Division of Plasma Physics, Denver, USA, 11-15 November 2013 IDM UID: KXGBBF

Physics basis and design of the ITER full-tungsten

divertor R. A. Pitts

ITER Organization, Plasma Operation Directorate, Cadarache, France

The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

With contributions gratefully acknowledged from:

B. Bazylev1, S. Carpentier-Chouchana, F. Escourbiac, J. P. Gunn2, T. Hirai, M. Kocan, A. Loarte, M. Lehnen, A. S. Kukushkin

1Karlsruhe Institute of Technology, IHM, Karlsruhe, Germany

2IRFM, CEA Cadarache, St. Paul Lez Durance, France



Content •  Introduction § Historical perspective on the ITER Organization

proposal to begin operation with a full-W divertor •  Physics basis for a W divertor § Expected lifetime, stationary and transient loads,

technology qualification •  Potential issues, design § W melting, accumulation, material evolution/erosion § Component tilting, shaping

•  Summary



Reminder: main divertor characteristics



Reminder: main divertor characteristics

Outer vertical target (JA)

Inner vertical target (EU)

Dome (RF)

Reflector plates (RF)

Pumping slot

Cassette body (EU)



Outer vertical target (JA)

Inner vertical target (EU)

Dome (RF)

Reflector plates (RF)

Pumping slot

Cassette body (EU)

Reminder: main divertor characteristics 54 divertor assemblies (~9 tonnes each)

4320 actively cooled heat flux elements

Bakeable to 350°C

NB: technology limit for steady state power handling on HHF areas: 10 MWm-2



H/He D/DT

BA

SELI

NE

New divertor strategy

H/He D/DT

~10 years

PRO

POSA

L

•  IO proposed in 2011 to eliminate first CFC/W divertor and begin operations with a full-W variant which should survive to the end of the first DT campaign



•  Spend ~2 years to: (a) develop a full-W divertor design

à Final Design Review, June 2013 (b) advance tungsten technology R&D

à excellent progress (c) assess the physics/operational risks of starting ITER

with a full-W divertor à IO, ITPA, dedicated experiments on tokamaks

IC recommendation (Nov. 2011)

•  Decision on whether or not to start with full-W to be made by late 2013

•  STAC-15 (14-16/10) and IC-13 meeting (20-21/11) •  IO requested STAC-15 to recommend to IC-13 that a

full W-divertor be used from the start of operations



Operational campaigns •  Analysis of what the full-W divertor will be required to

survive has been performed on the basis of the ITER Research Plan (approximate numbers and types of pulse) à grand total of ~25,000 pulses à Provide rough assessment of risks of starting with full-W and

guidance for sustained operation at high power in DT

•  Independent of schedule scenarios, always 3 main “operation phases” to first DT burn: He/H, DD, DT §  At most ~25% at the highest power/stored energy §  Perhaps ~20% in pure He §  More than 50% at relatively low flattop duration and low

power



Stationary and slow transient load cycles

~5000 cycles to 10 MWm-2 ~300 cycles to 20 MWm-2

•  Derived on the basis of the Research Plan and SOLPS divertor performance simulations

•  Defines technology qualification requirement

•  Peak heat fluxes to >20 MWm-2 during reattachment if λq is narrower than we think à most important in DT phase §  see R. J. Goldston, Y12.000001 for more on narrow λq



Divertor monoblock qualification •  Good results now from both JA and

EU Domestic Agencies – tungsten monoblocks tested in ITER Divertor Test Facility (electron beam) §  Technology meets design loading:

20 MWm-2 (300 cycles, 10 sec) §  Survived 1000 cycles at 20 MWm-2

KHI-MMC KHI-ALMT KHI-ATM

Ansaldo, Plansee

MHI-ALMT MHI-ALMT KHI-MMC

Mitsubishi and Kawasaki heavy Industry

IDTF, St. Pertersburg, RF



Steady state loading – “carbon free”

“Non-active” “Active”

A. S. Kukushkin, et al., EPS 2013 A. S. Kukushkin

10 MWm-2 limit for steady state power handling

•  New SOLPS simulation databases under construction for carbon-free divertor performance + impurity seeding §  Technology limits can be attained in He, H-mode at low Ip §  Simulations typically have λq ~4.5 mm (DT), λq ~ 3.5 mm (He)

PSOL = 60 MW

PSOL = 100 MW



Transient loads: ELMs

Ip (MA)

PIN (MW)

Wplasma (MJ)

Etransient (MJ)

λq||omp (m)

q⊥target (MJ m-2)

FHF (MJ m-2s-1/2)

7.5 40 75 6.4 à 8.0 0.01 1.39 à 1.74 77 à 123

•  Example: low power H-mode (likely in Helium)

NB: FHF,melt ~ 50 MJm-2s-1/2 for W



Transient loads: ELMs

Ip (MA)

PIN (MW)

Wplasma (MJ)

Etransient (MJ)

λq||omp (m)

q⊥target (MJ m-2)

FHF (MJ m-2s-1/2)

7.5 40 75 6.4 à 8.0 0.01 1.39 à 1.74 77 à 123

•  Example: low power H-mode (likely in Helium)

R. A. Pitts et al., J. Nucl. Mater. 438 (2012) S48 •  Uncontrolled ELMs in He/H phase unlikely to melt if ELM footprint broadens

•  Unless broadening very large at high current/power, or ELM-free regimes found, ELM control clearly required in the DT phases

•  Need to improve physics basis for broadening



ELMs and W accumulation

Lines to guide the eye only

ΔWELM (MJ)

Peak post ELM W radiation (MW)

DT high Q DT low Q He

Time dependent SOLPS for W source with ASTRA+STRAHL for core transport

•  ELM control with W divertor required as much for W impurity control as for material damage avoidance §  fELM > 20 – 30 Hz sufficient to

control W accumulation and post-ELM radiation spike for 7.5–15 MA D, DT H-modes

§  NB: fELM uncontrolled ~2-3 Hz §  He plasmas require higher fELM

due to higher sputtering and more unfavourable pedestal transport (if Ti,sep ~100 eV)



Transient loads: disruptions

Plasma stored

ene

rgy

[MJ]

No. of p

ulses

Flat to

p success rate

Stored

Ene

rgy achieve

rate

Major DisrupA

on ra

te

MD MiAgaAo

n success

rate

No. of m

iAgated MDs

MiAgated MD pa

rallel

energy flux at LCFS at

inne

r (ou

ter) ta

rget [M

J m

-‐2]

No. of u

nmiAgated MDs

Unm

iAgated MD pa

rallel

energy flux at LCFS at

inne

r (ou

ter)

target [M

J m-‐2]

He-‐H I 30

4200 0.8 0.5 0.2 0.75 252 6 (2) 84 49 (19)

90 0.5 0.2 0.75 252 18 (7) 84 146 (56)

He-‐H II 90

9800 0.9 0.5 0.12 0.85 450 18 (7) 79 146 (56)

150 0.5 0.12 0.85 450 30 (12) 79 243 (93)

DD 150

4950 0.9 0.5 0.08 0.92 164 30 (12) 14 243 (93)

210 0.5 0.08 0.92 164 43 (16) 14 341 (130)

Full Power DT QDT =10

210 5600 0.9

0.5 0.05 0.95 120 43 (16) 6 341 (130)

350 0.5 0.05 0.95 120 71 (27) 6 568 (216)

Plasma stored

ene

rgy

[MJ]

No. of p

ulses

Flat to

p success rate

Stored

Ene

rgy achieve

rate

DW VDE

rate

DW VDE

MiAgaAo

n success rate

No. of m

iAgated DW

VD

Es

MiAgated DW

VDE

pa

rallel ene

rgy flu

x at

LCFS at inn

er (o

uter)

target [M

J m-‐2]

No. of u

nmiAgated DW

VD

Es

Unm

iAgated DW

VDE

pa

rallel ene

rgy flu

x at

LCFS (o

uter baffl

e)

[MJ m

-‐2]

He-‐H I 30

4200 0.8 0.5 0.05 0.8 67 6 (2) 17 34

90 0.5 0.05 0.8 67 18 (7) 17 102

He-‐H II 90

9800 0.9 0.5 0.02 0.9 79 18 (7) 9 102

150 0.5 0.02 0.9 79 30 (12) 9 169

DD 150

4950 0.9 0.5 0.01 0.95 21 30 (12)



Disruption TQ transient loads

Rise times for TQ heat pulse: 1.5 – 3.0 ms (MD), 0.75 - 1.5 ms (VDE) Assumed broadening of heat flux footprint: 3 – 10λq|| Maximum in/out divertor asymmetry = 2

MD E||TQ (MJm-2) VDE E||TQ (MJm-2)

Unmitigated ~325 19 - 243 ~50 34 - 169 Mitigated ~1400 2 - 30 ~300 2 - 30

Unmitigated ~70 93 - 568



TQ transient heat flux factors MD FHF (MJm-2s-1/2) VDE FHF (MJm-2s-1/2)

Unmitigated ~325 26 - 328 ~50 65 - 321 Mitigated ~1400 3 - 41 ~300 4 - 57

Unmitigated ~70 126 – 768



Disruptions: transient heat flux variants VDE quenching on outer baffle or dome vicinity in limiter config. Current quench halo currents to baffles. Dome or baffle runaway electron impact

Mitigated downward VDE and mitigated+unmitigated in-place major disruptions

DINA simulations



Design: target tilting for steady loads •  All main plasma-facing components are tilted when

assembled onto the cassette body to ensure no leading edges from PFC to PFC

plasma flux



Design: target tilting for steady loads

0l0n

g axis+

80º

0.74º

0.5o

Til0ng axis

IVT OVT

Dome 0.7o

Til0ng axis 0.6

o

Til0ng axis

•  All main plasma-facing components are tilted when assembled onto the cassette body to ensure no leading edges from PFC to PFC



Design: shaping for transients

Monoblock chamfering to hide all edges in HHF areas

Outer baffle toroidal chamfering for VDE protection

Dome: no shaping à assess consequences

e.g. 0.5 mm



Melt estimates

*see e.g.: B. Bazylev, H. Wuerz, J. Nucl. Mater. 307-311 (2002) 69 B. Bazylev et al., J. Nucl. Mater. 390-391 (2009) 810

Monoblock surfaces: 2D (assuming edge protection) Mitigated and unmitigated Major Disruption Mitigated VDE Outer baffle and dome edges: 2D and 3D Unmitigated VDE (baffle, dome) Runaway electron impact (dome)

•  Use 2D and 3D MEMOS code* to assess PFC damage for surface and edge loading

•  Heating, melting, evaporation, vapour shielding, heat transport in liquid and solid, viscosity and melt motion (due to surface tension, external pressure and Lorentz forces)



Melt estimates: monoblocks

Assume simple chamfer, depth 0.5 mm, 3.5º total impact angle, monoblock width = 30 mm, gap = 0.5 mm ne = 1020 m-3, Te = 1 keV, M = 1 à p ~0.8 bar, cs ~5 x 105 ms-1 tdecay = 2trise, triangular heat pulse

Event E||TQ (MJm-2) Number trise (ms)

Unmit. MD 340, 570 6,6 1.5 - 3 Mit. MD 45, 70 120, 120 1.5 - 3 Mit. VDE 45, 70 12, 12 0.75 - 1.5

Plasma stream



Melt estimates: monoblocks •  E.g.: mitigated MD, E||TQ = 70 MJm-2, trise+tdecay = 4.5 ms •  2D MEMOS à tangential pressure force only, no currents

1 pulse 120 pulses

(cm) (cm)

•  For all ttotal < 9 ms, Tsurf > Tmelt, evaporation < 0.1 µm/pulse max. melt pool depth ~70 µm, melt motion 10-40 cms-1

•  Overall surface roughness (after 120 pulses) ~800 µm



Melt splashing

1: Melt splashing at monoblock edges: Macrocopic W droplet release and bridging of castellation gaps Cause: Rayleigh-Taylor instability or melt separation

2: Kelvin-Helmholtz instability Droplet breakaway

3: Disruption induced Eddy currents J×B forces perpendicular to melt layer at the TQ

1: vmelt too low for computed hmelt for melt layer separation vmelt too low and λR-T too large cf. hmelt for fast R-T instability

2: Density of impacting plasma too low to generate K-H instability

J×B

B J

hmelt vmelt •  3 main areas of concern:

•  Use MEMOS calculations (hmelt, τresolid, etc.) to conclude:



Melt estimates: RE dome impact

Dome tile Dome

tile

RE 2 mm

~10 cm

•  RE damage a potential problem in all phases of operation

•  Use ENDEP and 2D MEMOS for RE on inter-cassette dome misalignment (max ~3 mm)

•  Difficulty is specifying §  Total wetted area for RE impact §  Total energy carried by RE (e.g.

conversion of poloidal magnetic field energy to RE kinetic energy)

§  Energy deposition time



Melt estimates: RE dome impact

Y

ENDEP MEMOS

RE Dome

tile

T > 7000 K will

convert into

plasma

Melted W

•  Example: fast RE loss: tloss = 100 µs WRE = 20 MJ, ERE = 12.5 MeV

•  Very serious damage to be expected à no gain from shaping. NB: problem just as bad for main Be first wall



Material issues (active phases) •  The most outstanding long term “risk” for high power

operation on a W-divertor in ITER §  Will transient melting really occur according to

calculations? à recent JET melting experiment §  High fluence, high power operation (including sub-melting

threshold transients) on melt damaged W §  Erosion/evolution of recrystallized surfaces under plasma

impact (cracking, roughening) §  Surface modification due to He plasma impact (e.g. nano-

structures (W-fuzz)) §  Possible Be-W alloying



Example: operation on damaged W •  Create melt-damaged

area (Pilot-PSI) using high Tsurf and small transients to take surface above Tmelt

•  Expose to high fluence plasma + transients (Magnum-PSI) §  Study surface response with

high resolution IR §  Examine post-mortem with

microscopy and metallurgy

Courtesy of G. De Temmerman, DIFFER

After 72 shots, ~4100 ELM-like transients



Material issues (non-active phases) •  For non-active phases risk of starting full-W appear to

be low* with some caveats: § Concern that pure He operation at lower power (e.g.

H-modes at 7.5 MA) may lead to enhanced erosion through bubble formation driving reduced surface thermal conductivity à more R&D required here

§  Adequate disruption mitigation, ELM control and diagnostics must be in place at appropriate stages in the experimental campaigns

*see R. A. Pitts et al., J. Nucl. Mater. 438 (2013) S48

•  A vigorous long term R&D programme required to study W material issues during ITER construction



Concluding remarks •  After a crash programme beginning in late 2011:

§  A mature full-W divertor design is in place with key supporting analysis nearly complete

§  Technology in all the supplying Domestic Agencies meets and even largely exceeds the cyclic load specs



Concluding remarks •  Key identified risks from the plasma operation side are

W melting, material surface evolution and core plasma contamination § Melt splashing not expected, but surface topology

modifications likely à possibly on mm scales §  ELM control required for PFC damage avoidance (high

power) and W accumulation control (all H-modes) §  Long term material surface evolution under high power,

high fluence plasma impact in the presence of transients still an area requiring much R&D à was always the case, even in previous ITER divertor strategy



Concluding remarks •  Recommendation from ITER STAC after their 15th

meeting (14-16 October 2013): §  The ITER Council adopt the IO proposal to implement in

the Baseline the option of choosing W targets for the first ITER divertor

•  ITER Council to decide at the next meeting (IC-13, 20-21 November 2013)



Reserve material



Outer target loading profiles B2-Eirene, A. Kukushkin

~10 MWm-2"

All-metal, PSOL = 100 MW, 0.4% Ne Carbon, PSOL = 100 MW

•  Baseline” cases retrieved for Ne seeded no carbon operating point at qpk ~10 MWm-2 §  But often large differences in contributions to target loads à

Ne and C do not radiate from same location

Documents

Physics basis and design of the ITER full-tungsten divertor1 @2013, ITER Organization 55th Annual Meeting of APS Division of Plasma Physics, Denver, USA, 11-15 November 2013 IDM UID: