Key Issues in Plasma-Wall Interactions for ITER The European Approach V. Philipps, J. Roth, A. Loarte With Contributions G.F MatthewsH.G.EsserG. Federici.

Key Issues in Plasma-Wall Interactions for ITER

The European Approach

V. Philipps, J. Roth, A. Loarte

With Contributions

G.F Matthews H.G.Esser G. Federici

J.P.Coad U.Samm M. Mayer

J.Strachan P.Wienhold P.Andrew

M.Stamp A. Kirschner G. Pautasso

M. Rubel W. Jacob

Key Issues in Plasma-Wall Interactions for ITER

The European Approach

V. Philipps, J. Roth, A. Loarte

With Contributions

G.F Matthews H.G.Esser G. Federici

J.P.Coad U.Samm M. Mayer

J.Strachan P.Wienhold P.Andrew

M.Stamp A. Kirschner G. Pautasso

M. Rubel W. Jacob

EU-PWI-Task Force

EPS 2003, Petersburg

Magnetic Confinement Fusion is ready to build a first machine delivering 500 MW fusion power : ITER

Controlled Thermonuclear Fusion has the potential to open a new primary energy source to mankind

The fuel (deuterium and lithium) is cheap and worldwide accessible

– This is also a contribution to a peaceful world

EU-PWI-Task Force


Magnetic Confinement Fusion

EU-PWI-Task Force


ITER – a worldwide undertaking

Europe, Russia, Japan, Canada, USA, China, S.-Korea

goals• 500 MW fusion power, Q=10 with burn time 7 min

• Quasi-steady-state plasma operation,with Q=5 and 'Hybrid' scenarios with pulse length up to 30 min

• Integration of physics and technology (Tritium, breeder blanket, super conductors, heating)

Four ITER sites offered

Four ITER candidate sites

Magnetic Confinement Fusion

V. Mukhovatov, I.3.3, Wed

JET has achieved simultaneously the essential dimensionless ITER parameters

Confinement

Pressure

Density

Purity

RadiationShaping

Pulse duration

From JET to ITEREU-PWI-Task Force


JET ITER

Scaling: Plasma performance

JET ITER

EU-PWI-Task Force

PULSE LENGTH (S)

STORED ENERGY (MJ)

INPUT ENERGY/ SHOT (MJ)

DIVERTOR PARTICLE FLUENCE/ SHOT

JET 40 10 40 1x 1024

ITER 400 350 50000 4 x 1027

x10 x35 x 1000 x 4000

Challenge to Technology and Plasma Wall Interaction

ELMs and disruptions

Lifetime and T-retention

From JET to ITER

EU-PWI-Task Force


• Control of MHD modes (NTM)

high plasma pressure

-particle heating

• Current drive efficiencies

• Control of steady state heat load

• Control of transient heat loads (ELMs and Disruptions)

• Lifetime of plasma facing components

• Long term Tritium inventory limit (350g)

ITER is an experiment to analyse these questions.

Remaining crucial issues

• All present data from carbon devices indicate a long term fuel (T) retention which would be unacceptable for ITER

JET T experience

Long term fuel (T) retention

JET (T) 10%

TFTR (T) 13%

TEXTOR (D) 8%

Similar observations in Tore-Supra and various devices

Equivalent ITER T limit (350g) would be reached in less than 50 shots

10% long term retention

EU-PWI-Task Force


Fuel retention: present database

Of injected fuel

T. Loarer P-1.161, Mon

EU-PWI-Task Force


ITER

700m2 Be first wall Low Z Oxygen getter

100m2 Tungsten low erosion

50 m2 Graphite CFC no melting

ITER wall material Choice

A European Task Force on Plasma Wall Interaction has been formed to focus the EU- PWI research on the critical questions of Tritium retention and Wall Lifetime.

A European Task Force on Plasma Wall Interaction has been formed to focus the EU- PWI research on the critical questions of Tritium retention and Wall Lifetime.

ITER has different first wall materials

Simple extrapolation from present full carbon devices is not possible.

Must be based on physics understanding.

Long term tritium retention

EU PWI Task Force Strategies

Understand (better) the mechanism of fuel retention in present devices

• Improve predictions for ITER

• Develop Tritium control techniques

A

Develop Tritium removal techniques that are applicable for ITER

B

Develop a full metal Tokamak scenario

C

EU-PWI-Task Force


EU PWI Strategies

EU-PWI TF structure

• Coordinated experiments and data analysis in JET (Task Force E & FT) and

EU Fusion associations

• Accompanying Technology Programme

• Contact persons in each association

EU-PWI-Task Force


Diffusion along pores

Implantation(saturates)

D+ D+

C

Erosion area Deposition area

Remote area

• Tritium is retained by co-deposition with carbon, on the plasma facing sides or on remote areas

Understanding of T-codeposition is understanding of

where and how carbon is eroded and

how carbon migrates globally and locally

Fuel retention: Understanding

Co-ordinated research in Tokamaks and lab experiments in PWI-TF

EU-PWI-Task Force


outer

Erosion and Deposition in Divertor (1)

P. Coad et al, PSI GIFUJET gas JET gas box, box, 5750 shots

Ero

-dep

osi

tio

n (m

)

JT-60 4300 shots

inner

Inner Divertor tileEro

-dep

osi

tio

n (m

)

Outer Divertor tile

Y. Gotoh et al , PSI GIFU

Inner Divertor is deposition dominated in all devices

The outer divertor can be erosion or deposition dominated

Depending on ?

In/out asymmetry of Divertor Conditions

Differences in SOL || Flows

Influence of temperature

Divertor Geometry

Adressed in PWI-TF

EU-PWI-Task Force


ASDEX Upgrade, PSI 2002, V. Rohde

Inner Divertor

Outer

V.Rohde P-1.154, Mon


1

4

3

6

5

• Beryllium is deposited on the plasma facing areas, no transport to shadowed regions

• Carbon and deuterium is mainly transported to shadowed areas

Transport to remote areas is specific to carbon

EU-PWI-Task Force

3

4

C: Be = 10:1

CarbonBeryllium G. Matthews P-3.198, Thurs

Carbon


EU-PWI-Task Force


Local geometry determines the C-deposition on the louver entrance

QMB and sticking monitors (M. Mayer O-2.6A, Tues) show that the

carbon deposition is mainly line of sight of the place of origin

Quartz monitor (QMB)

1 2 3

3

0,0

0,2

0,4

0,6

0,8

1,0

1,2

Configuration

C-d

epo

siti

on

(n

m/s

)

1

22

3


EU-PWI-Task Force


13CH4 tracer injection in TEXTOR

LCFS13CH4

Plasma

P. Wienhold. A. Kirschner, PSI 2000

Modelling of erosion and redeposition

A. Kirschner et al

240230 250 260220

-170

-160

-150

-140

-130

-180

RC [cm]

ZC [cm]

C0 Density

• With standard assumptions (2% erosion yield, „TRIM“ sticking of redeposited species):

- modelled C-fluxes to the louvres much too low (JET) and - locally redeposited carbon (TEXTOR) much too low

• Good matching of Be transport

JET MKIIA

EU-PWI-Task Force


Assumptions: carbon atoms eroded in a first step can be re-eroded with higher yields after re-deposition

Enhanced movement of carbon along surfaces to shadowed areas

Assumptions: carbon atoms eroded in a first step can be re-eroded with higher yields after re-deposition

Enhanced movement of carbon along surfaces to shadowed areas

Trim sticking for ions, zero sticking for CxHy

8% re-erosion of re-deposited carbon species

Understanding of carbon transport

Physics of sticking and re-erosion is the key to understand carbon long range transport

Substrate chemical erosion Yield 2 - 3%

D CH4

DCH4CH+

Shadowedareas

Graphite

Tungsten

Standard assumptions

Carbon deposition: 5% of C-erosion flux 0.7 gT retention / ITER shot

Enhanced re-erosion

Carbon deposition: 14% of C-erosion flux 2 gT retention / ITER shot

Eroded carbon can escape towards the dome and dome pumping ducts

T-removal should be considered there

EU-PWI-Task Force


Modelling for ITER Divertor

A. Kirschner P-3.196, Thur

Modelling

EU-PWI-Task Force


Erosion redeposition in divertor: summary

• Inner divertor deposition dominated always

• No unique behaviour of outer divertor

• Long range transport is specific of carbon

• Main chamber erosion dominated area in general (with local or global material redistribution)

• The material deposited in the divertor is mainly from main chamber erosion (mostly C at present, Be in ITER)

JET: material balance, divertor Be deposition

AUG: material balance and tungsten divertor experience

DIII: spectroscopic analysis

Main chamber ion PWI is significant and underestimated in the past

”long tails” in SOL ne & Te seen in many experiments

ASDEX Upgrade

Te

ne

• larger divertor closure moderate decrease of neutral pressure in main chamber

• minimum main chamber pressure set by main chamber ion plasma wall contact

• main chamber contact determined largely by ELMs?

J. Neuhauser et al.

SOL profilesNeutral Pressure measurements

EU-PWI-Task Force


Main chamber Plasma Wall Interaction (1)

W.Fundamenski O-4.3C, Fri

A.Herrmann P-1.155, Mon

B. Lipschultz P-3.197, Thurs

A.Kallenbach P-1.159, Mon

Main chamber Plasma Interaction is main topic in TF work

EU-PWI-Task Force


Main chamber erosion• Absolute main chamber PW interaction

Divertor / First Wall fluxes: 12 (JET), 10 (AUG), 50 (ITER modelling)

• Erosion Mechanisms Ions versus Neutrals

• Erosion Location HFS versus LFS

Material migration

• Measurements and understanding of SOL Flows Modelling based on ExB and Bxgrad B drifts

underestimates measured flows

Important to predict ITER outer divertor behaviour

Main chamber Plasma Wall Interaction (2)

EU-PWI-Task Force


The ITER Be-first wall will reduce the Carbon deposition and associated T-retention

1. No C-flux into the divertor, but a similar Be-flux

[present modelling: 6 g Be/shot, better quantification needed]

2. Be is not transported to remote areas

3. Be-layers on the plasma facing sides of the divertor contain less T and are easier to access for cleaning

4. Chemical sputtering of the underlying C-substrate in the inner is reduced/suppressed

T-retention: Extrapolation to ITER

Be transport and influence of Be deposition on carbon erosion & transport are key questions for ITER ( PWI TF)

EU-PWI-Task Force


Chemical erosion completely suppressed by adding 0.1 % Be to the plasma

Questions to address

• Thermal stability of Be layer during transient heat pulses

• Be erosion-deposition in the outer divertor

PISCES

R. Doerner P-2.162, Tues

Beryllium experiments in PISCES

EU-PWI-Task Force


Fuel removal and control

Control of fuel retention and fuel removal will be essential in any wall material scenario and needs more attention in

present research (major topic of PWI-TF work)

Fuel Removal

• Isotope exchange on PFC side

• Thermal desorption on PFC side

• Oxygen venting remote areas?

• Scavenger techniques ??

and Fuel Control

• Temperature tailoring

• Carbon traps

• Divertor geometry …

Work in plasma simulators

+

Dedicated lab experiments

+

Tokamak research

Needs detailed understanding of the involved physics

EU-PWI-Task Force


Full metal wall: wall lifetime

Steady state erosion

• Low erosion materials

• High local re-deposition

• Lost material replaced from main chamber

D+

99.8 0.002

Plasma

Divertor Strike zones

If the T-retention problem cannot be solved a full metal first wall concept is needed based on metals with low hydrogen retention

Plasma compatibilityWall Lifetime

Transient events

The main concern with metal walls is the lifetime due to melt layer erosion in transient heat loads (ELMs and Disruptions)

Particle flux

Stored energy

EU-PWI-Task ForceLifetime ELMs & Disruptions (1)

• In ELMs or disruptions part (<10%) or all of the plasma-stored-energy is lost on a short-time scale to the walls

Material limits T < 2300o (C), < 3400o (W)

T energy/ area/ sqrt(time)

Wetted area Duration of ELMsEnergy on target plates

20 (40) MJ m-2 s -1/2

Divertor wetted area during ELM similar to between ELMs

some ELM energy can reach the main chamber

T. Eich, A. Herrmann, this meeting

EU-PWI-Task Force

Duration of Divertor ELM Energy Pulse well correlated with II B Ion Transport 220s for ITER

Elm / in between Elm

Lifetime ELMs & Disruptions (2)

EU-PWI-Task Force


number of ELMs for 1000 ITER pulses 4 x 105

Predictions for ITER divertor target lifetime

ITER Predictions

ELMs are marginally acceptable but behaviour for CFC and W not much different at moderate power densities

Lifetime: ELMS & Disruptions (3)

G. Federici et al.

Disruption Energy deposition

(Thermal Quench) occurs over a

large divertor area (AUG & JET)

Fraction of energy deposited in

Divertor: 1 (AUG), ~ 0.2 (JET)

ASDEX Upgrade

Poloidal Distance (m)

V. Riccardo, I.3.5. Wed

G. Pautasso P-1.135, Mon

P. Andrew P-1.54, Mon

EU-PWI-Task Force


Disruptions: Present ITER specifications

All disruption energy is lost at the strike zones

with narrow deposition (broadening of 3)

half of the melt layer is lost per event

Lifetime: ELMS &Disruptions (4)

EU-PWI-Task ForceD

isru

pti

on

en

erg

y d

en

sit

y,

MJ

/m2

SOL broadening, lambdaplate

/lambdamid

Current ITER specifications

G. Federici, ITER JWS Garching, 7 Oct. 2002

Absence of melting for W

lambdamid

=5 mm

Dis

rup

tio

n e

ne

rgy

de

ns

ity

(M

J)

1

0.5

0.1

0.1

1

10

100

0 5 10 15 20 25 30

Fraction of energy to divertor

• Spatial and Time Evolution of thermal Energy Flux

• Dependence on Disruption Type

• Energy Balance during Current

Quench (Halo Currents)

PWI TF issues

Disruptions energy deposition and mitigation are key issues for the ITER material selection

SOL Broadening

Lifetime: ELMS &Disruptions (5)

• Development of disruption mitigation techniques

• Power exhaust on irregular, molten surfaces

EU-PWI-Task Force

Conclusions (1)

Predictions for long term tritium retention for ITER are critical, but are from full carbon machines. An integrated approach and understanding of

• Where and how impurities are produced in the main chamber

• How they are transported towards the divertor

• How the material is transported inside the divertor

• Influence of Be deposition on carbon erosion and transport

is necessary to predict the T retention under the ITER wall material conditions.

EU-PWI-Task Force

Conclusions (2)

In parallel co-ordinated work is necessary on

• In situ control of T retention

• Fuel (T) removal which can be employed under ITER conditions

• Disruption power deposition characteristics

• Disruption mitigation

More Tokamak experience is needed for metal wall conditions

• High Z first wall, tokamak behaviour under non- carbon wall conditions

• Be first wall with carbon and tungsten in the divertor

R.Neu P-1.123, Mon

EU-PWI-Task Force

Additional remarks

• A graphite and a tungsten divertor should be prepared in parallel.

• Possibilities to measure the fuel retention in the non activated phase of ITER are needed.

• The possibility to change ITER in a later state from a low to a high Z first wall should be evaluated.

• ITER needs flexibility to adopt Tritium control and removal techniques that have to be developed in parallel and tested in present devices.

Key Issues in Plasma-Wall Interactions for ITER The European Approach V. Philipps, J. Roth, A. Loarte With Contributions G.F MatthewsH.G.EsserG. Federici.

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