iter-scientific-status elms and disruptions

8/3/2019 ITER-Scientific-Status ELMs and Disruptions

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Page 152

nd

APS Division of Plasma Physics Meeting , Chicago, Illinois, USA

Agenda Item xxx

ITER Scientific Status and RequiredR&D

Alberto Loarte

Plasma Operations Directorate

ITER Organization

Acknowledgment : contributions from IO staff, Domestic Agencies, ITPA, EU-

Task Forces, US-BPO, and ITER Members Fusion Research Institutions


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Page 252

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1. Introduction2. Open R&D issues with major influence on designs for Baseline

Heat Loads on PFCs ELM Heat Fluxes and ELM Control Schemes Disruptions Loads and Disruption Mitigation (heat, forces and runaways)

3. ITER Scenarios and Open R&D Issues H-mode access (incl. Ip ramp), control of H-mode access and H = 1 sustainment Helium H-modes Fuelling of H-mode Plasmas Control of plasma during transients NTM control, RWM control, etc.

4. Summary and Conclusions

Outline of Talk


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ITER Mission and Design (I)

ITER Mission : To demonstrate the scientific and technological

feasibility of fusion energy for peaceful purposes

ITER fusion performance goals dominated plasmas (P/Padd 2 QDT 10) with Pfusion = 500 MW

Inductive operation with 300-500 s burn time Plasma performance H-mode scaling H98 ~ 1 and /nGW ~ 0.95 achieved in

present tokamaks in high Type I ELMy H-mode regimes

Long pulse operation (~ 1000s) with P/Padd 1QDT 5 with Pfusion 350 MW Most plasma current self-driven (bootstrap) + externally driven Plasma performance enhanced confinement regimes : hybrid scenarios or H-

modes with Internal Transport Barriers

Definition of plasma regime that meets ITER requirements subject of R&D


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ITER Mission and Design (II)

Machine mass: 23350 t (Cryostat + VV + Magnets)

Magnets

Vacuum Vessel

Blanket

Divertor

Cryostat

Diagnostics and H&CD systems (33 MW NNBI, 20 MW ICRH, 20 MW ECRH)


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Page 552nd APS Division of Plasma Physics Meeting , Chicago, Illinois, USA

ITER Design Review 2007- 2009 Re-assessment of ITERDesign Capabilities to Achieve Projects Mission Increased Current Requirements for PF coils and CS Force Limit

Modification of PF6 (lower divertor-coil) location and Current Requirement Divertor Geometry modified for lower li operation at 15 MA (high Pped) Shaped & Detachable First Wall II Plasma Fluxes & Replaceability In-Vessel Coils for Vertical Stability Control Need for ELM Control Identified In-vessel ELM Control Coils, Pellet Pacing,

(decision still open)

Main Focus of Work 2009-2010 Definition of ITER Research plan (First Plasma DT) in Conjunction with

Phased Installation & Commissioning of Systems

Further Studies of ITER Scenarios/Plasma Conditions Input to DetailedDesigns & Required Physics R&D

Progress in Detailed Design of Components/Systems (FW, H&CD, ) Assessment of Transient Load Control Requirements (ELMs and Disruptions)

and Scheme Designs for Inclusion in ITER Baseline

Review of Activities 2007-2010


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ITER Experimental Schedule to DT

2019 2020 2021 2022 2023 2024 2025

First

Plasma

ITER Commissioning and Operations

DD & Trace DT Operations

Full DT

First

Plasma

H & He Operations

Coil Commissioning (&H plasma possible)

Shutdown

Commission

Shutdown

Install In-

Vessel

Equipment,

ECRH &

Diagnostics

Install Blanket,

Divertor, NBI 1+2,

ECRH, ICRH

Diagnostics,TBMs

2026

Hydrogen Operations

Q=10 short pulse

All H&CD Fully Commissioned

Tritium Plant Ready for Nuclear Operation

Tritium Plant Full DT Throughput

H & He Operations

Pre-Nuclear Shutdown

2027

Install

Diagnostics

TBMs

Commission

Neutron Diagnostic CalibrationDivertor Change

Q=10

Short PulseFull DT


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ITER Research Plan - Major Elements

H/ He Campaign I: March 2022 - January 2023 System commissioning with plasma H&CD short pulse commissioning to ~70MW input power 15MA/ 5.3T technical demonstration

H/ He Campaign II: November 2023 - May 2025 H&CD commissioning to long pulse Disruption loads completed/ disruption mitigation implemented ELM control commissioned in helium H-modes

D/ DT Campaign: May 2026 - August 2027 Commissioning of Tritium Plant with tritium Commissioning of tungsten divertor in H/ He plasmas Development of H-mode scenarios in deuterium Trace tritium experiments begin in January 2027 Full DT experiments begin in March 2027 Attempt at Q=10 short pulse in August 2027


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ITER controller with free-boundarycoupled to transport

All coil currents remain within limits Voltage waveforms realizable with

new power supply design

Evolution of density and Zeffprescribed

Access to H-mode assumed with52MW auxiliary heating

Focus on H-mode performanceat flat top rather than H-mode

access QDT=10 performance and burn

duration meet ITERs mission

ITER QDT 10 Scenario

T. Casper IAEA 2010


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ITER QDT 5 Long Pulse Mission

C. Kessel IAEA 2010

Ip

= 12.5 MA

IBS = 3 MA

INB = 1.4 MA

PNB = 33 MW

PIC = 20 MW

Palpha = 82 MW

Prad,core = 42.5 MWQ = 7.7

li(3) = 0.94

n/nGr= 0.88

N = 2.15

H98 = 1.25

Zeff= 2.0fNICD = 0.4

Tped = 4.5 keV

n(0)/ = 1.07

tburn > 1000 s


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Top

Inboard

Outboard

BM #1-6

Central columnHFS start-upToroidal & poloidalshaping

BM #7-10

Secondary divertorregionToroidal & poloidalshaping

BM #11-18OutboardLFS start-up/ramp-downToroidal shaping

All Be First Wall Panels shapedShape & Power Handling ( 2 or 5 MWm-2)result of (on-going) optimization between steady

loads and transients

ITER First Wall Design

R. Mitteau


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In-vessel coils

VS and ELM control coils (also RWM) Successful PDR in October 2010 Scientific case for VS coils universally supported (Design and Conductor R&D on-going) Decision on Adoption of ELM coils into Baseline to be taken by June 2012 at the latest

strengthen Scientific Case or Develop Alternative ELM control methods

Design, Integration and R&D to continue for all in-vessel coil systems (FDR ~ Feb 2012)

Upper VS

coil

Lower VS

coil

ELM

coils

VS Coils Normal

Operation

Number 2 coils - 4 turnseach

Maximum

current

(pulsed)

240 kAt/coil

Voltage 2.3 kV

ELM CoilsNumber 27 coils - 6

turns each

Maximum

current

15 kA (+ 90

kAt/coil)

Voltage 230 V


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Open R&D : Near SOL heat fluxes

No physics basis for inter-ELM near-SOL power channel and scaling to ITER q ~ 5 mm for ITER from SOLPS modelling and stability arguments but

could it be much lower?

New results indicate strong negative Ip scaling very narrow width for ITER Physics of qII ? Potential issues for baseline steady state heat flux

handling and divertor conditions (sweeping, He pumping )

DIII-D, Makowski et al. PSI 2010 NSTX, Gray et al. PSI 2010

Influence of RMP coils on near SOL power flux scaling ?

DIII-D M. Jakubowski- NF 09


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Page 13

52nd APS Division of Plasma Physics Meeting , Chicago, Illinois, USA

Progress in understanding divertor target heat loads ELM wetted area increases with DWELM ELM filaments Good news for Ip range possible without ELM control in ITER Small influence for 15 MA requirements if AELM = Abet-ELM for small WELM Physics of AELM(WELM) needs to be understood for extrapolation to ITER

JET, T. Eich PSI 2010 DIII-D, M. Jakubowski NF 2009

Open R&D : ELM SOL heat fluxes

Understanding of First Wall ELM loads for large & small ELMs + consistency withdivertor observations needed (WELM control limit could be set by FW)


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Pellet local edge over-pressure ELM triggering (Huysmans, THS/7-1)Experiments : Up to ~ 5 x fELMuncont increase in DIII-D (fELMcont ~ 1.8 fpellets)

with ~ 10% E decrease (Baylor EPS10)DIII-D Baylor -EPS10

ITER requirement of ~ 30 fELMuncontrolled and effects on Wplasma need to be assessedAdditional qELM from pellet particles expulsion by ELM needs to be understood

DIII-D Baylor -EPS10

Open R&D : ELM pacing by pellets


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ITER in-vessel coils with DIII-D guideline Icoilmax = 90 kAt (20% margin) &one power supply/coil for flexible perturbation alignment

n = 4 |br|/BT,0 ~ 6.6 10-4

O. Schmitz PSI 10

fcoil 5 Hz to allow perturbation rotation > 1 Hz smoothing of possible hotspots or localised erosion regions without PFC thermal cycling

20% Icoil margin provides system resilience to coil failure design criterionmet for Ip 14.5 MA with up to 3 failed coils in rotating mode

Open R&D : ELM suppression by RMP (I)


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Physics basis for ELM suppression in development extrapolation uncertain

Magnitude of |br/BT,0| for ELM suppression in ITER sufficient penetration ofresonant perturbation in ITER edge plasma?

Effect on density, fuelling, radiative divertor : low fuelling efficiency byrecycling in ITER controlled by pellet fuelling and no NBI fuelling Lower |br/BT,0| required in ITER & less effect on ? Avoidance of ELMs following pellet injection ? ELM suppression at /nGW ~ 0.9 & *ped


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Maximum allowable burst of gas into VV torecover operational conditions without

significant operation delay is limited

Gas for MGI ITER system limit

(kPa*m3)

D2 50

He 40

Ne 100

Ar 100 (


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~ 0.3 kPa*m3 of Ne needed to re-radiate plasma thermal energyreduces CQ to ~ 75 ms

Reasonable window of 0.3 -10kPa*m3 to mitigate thermal loadswithout excessive forces on the in-vessel components

Runaway avalanche suppressionby collisional damping probablyonly viable if n < 0.5 nRosenbluth

Open R&D : Disruption Mitigation Thermal Loads & Forces (II)

S. Putvinski IAEA 2010


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Large magnetic perturbations and secondary disruptions can be producedby dense gas jets injected repetitively in the CQ plasma

Required gas pressure ~ 1 atm, gas amount ~1 kPa*m3, 5 jetsstaggered in time by 5 ms --> Total amount of gas can be 10 times

less then for collisional damping!

Test of schemes of this type or other viable alternatives for mitigationof runaway loads is urgently required for ITER

Dense and resistive gas jetcontracts current channel

Modeling of RE suppression

Open R&D : Runaway Mitigation

S. Putvinski IAEA 2010


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Open R&D : H-mode Access

Power requirements for H-mode access in ITER evaluated in terms of globalscaling law

Large scatter in part experimental variability but also hidden parameters edge parameters and study of experiments with systematic deviations(X-point height, input torque, )

Study H-mode access for ITER-specific scenario requirements (in Ip ramps)Y. Martin, et al., Jour. Phys. Conf. (2008)


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Similar effect seen in several devices and can more than double L-Htransition power for similar global parameters

(Zx-Zbot)/a ~ 0.5 (ITER), 0.3 (JET), 0.4 (DIII-D)Unclear driving change in local parameters and PL-H if neutral escape thennot an issue for ITER good test for H-mode models

Coordinated ITPA experiments dependences in local & global parametersacross devices

JET-Andrew

DIII-D-GohilJET-Andrew

Open R&D : H-mode Access and X-point Height


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ITER QDT = 10 scenarios are designed with H-mode phases at Ip


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Open R&D : Control of H-mode access exit from H ~ 1

Ip = 15 MA - DINA ITER V. Lukash & Y. Gribov

Access and exit to H ~ 1 strongly dependent on P behaviour aroundtransition

P strongly dependent on pedestal and core plasma build-up/build-down afterL-H/following H-L transition (in particular on )

Experiments to characterize edge/core evolution around L-H/H-L transition

(ITPA) and burn-simulation experiments required to assess expected behaviour

in ITER and to develop control schemes for ITER


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Open R&D : H ~ 1 sustainment in ITER (I)

Stationary H ~ 1 can require up to Pinput > PL-H for ITER QDT=10. > 1 may depend on factors () which do not affect PL-H

ASDEX-Upgrade-Ryter-H-mode WS2007

ITER QDT =10, 500 MW Padd=50 MW, P=100 MW, Pradcore=50 MW (1.3)

JET-Saibene PPCF 2002


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Influence of edge/divertor radiation as required for acceptable qdiv on

confinement is a major issue to address for ITER

Necessary to understand to which level ELM dynamics, edge power flow, H-mode hysteresis, etc., affects HH ~1 sustainment in ITER

JET-Sartori H-mode workshop 2009 C-Mod-Hughes-IAEA10

Open R&D : H ~ 1 sustainment in ITER (II)


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Open R&D : Characterisation of He Type I ELMy H-modes

He Type I ELMy H-modes are key to development of ITER Research Plan :H-mode access & H-mode confinement at ITER scale and development of

ELM control techniques Assessment of key issues for ITER neededbeyond L-H threshold power requirements

Access to Type I ELMy H-mode, ELM characteristics, He H-mode fuelling, Influence of H on He for H-mode and Type I ELMy H-modes required to

assess viability of pellet pacing in He plasmasASDEX-Scarabosio-EPS09

1 MA

0.6 MA


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Edge density and plasma fuelling in ITER expected to be different frompresent devices if ionisation and diffusion dominate edge transport :

edge plasma dense and hot inefficient fuelling of pedestal plasmaby neutrals

density pedestal width determined by pedped = Dp (nped-nsep)/wn

DT_s ~ 6 1021s-1

ITER-B2-Eirene

Kukushkin

Open R&D : Fuelling of ITER H-modes (I)


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Neutral fuelling of plasmas in conditions of edge neutral opacity approachingthose of ITER role of sources versus transport in pedestal fuelling

Assessment of fuelling by neutrals in ITER-like conditions required tounderstand reliance on pellet fuelling for all phases of discharges and

fuelling of He plasmas

Nunes H-mode workshop 09JET 2MA-Kallenbach PPCF04

ion/Wn ~ 1/3-1/2

Open R&D : Fuelling of ITER H-modes (II)

ion/Wn ~ 1/3-1/2


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Open R&D : Fuelling of ITER H-modes (III)

Main plasma fuelling of ITER for high QDT regimes based on pellet injectionPellet size (50-90 mm3) & speed (300-500 ms-1) from modelling/experimentsUncertainties remain :

Ablation typically > 0.95 pellet penetration by drift understandingof drift scaling with device size and nped, Tped, etc. required

Loss of pellet-injected particles by following ELMs needs quantification A. Polevoi NF05

B. Pegouri EPS09


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Most ITER Baseline systems are in procurement or detailed design phases R&D is needed in some areas to take decisions on few remaining systems or

detailed design choices (timescale 1.5 years from now)

ELM control schemes Disruption Mitigation schemes with emphasis on runaway suppression (or soft landing if

needed)

Detailed design of First Wall Panel Development of ITER operational scenarios (non-active to DT) requires R&D

to determine plasma behaviour and use of baseline systems for its control

H-mode access/sustainment (including Ip ramp-up/down phases) Access to H ~ 1 from low confinement H-mode and control of P (through ) Sustainment of H ~ 1 and relation to ELM control requirements He H-mode plasmas characterisation and control of ELMs Fuelling of ITER high Ip H-modes : sources vs. pinch and pellet fuelling

Plasma control during confinement transients MHD control (NTM, sawteeth, RWM, ) Continued R&D support by fusion community required to guide outstanding

decisions on ITER Baseline systems/detailed designs and for the definition of

realizable ITER operational scenarios

Conclusions


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ITER Reference Plasma ParametersTable shows nominal plasmas parameters for ITER scenarios


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CS primarily ohmic current drivebut can be used to move plasmaaway from inside wall

VS1 (PF2,PF3) and (PF4,PF5)differential currents for stability

control

VS2 can be used for control notin baseline VS3 new internal coils closely

coupled to plasma for fast response

Disturbance control

Reduce effects of noise incontrol

ITER PF System


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ITER H-mode Power Threshold

The latest H-mode threshold power scaling for deuterium plasmas:

The isotope dependence based on JET results in H, D, and DTindicates that P

thresh 1/A for hydrogen isotopes

Note: within the ITER formalism, input power normally corrected forcore radiation fraction of ~30%

(Y Martin, HMW-2008)

half-field/ half current H-mode development

Full-field/ full current H-mode development

No H-mode access in D for full Q=10 simulation No H-mode access in H at full field

H-mode access path in DT needs 40MW

Q=10

Possible helium H-mode access


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Effects of torque input seen in several devices but effects vary from device todevice and within device for different conditions

If input torque/rotation effects important scaling law probably overestimatesITER requirements (if ITER rotation is low)Systematic/Coordinated assessment in tokamaks with well diagnosed edge

rotation and n-T, etc., required to make progress for ITER

DIII-D-Gohil

JET-Andrew

ICRH

NBI

C-mod-Rice

Open R&D : H-mode Access and Torque Input


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PL-H increases strongly below a given densityUnderstanding of low density limit and predictions for ITER are very uncertainMajor issue is whether high limit in C-Mod is relevant to ITER or notalthough there seems to be a favourable machine size scaling

Factors affecting L-H transition : Low ne limit

Martin JPCS09 + C-Mod-SnipesC-Mod-Snipes


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Typically, experiment dependent and thus difficult to evaluate in ITERMore effort in developing techniques compatible with ITER operation pellet

injection, current ramps (down), X-point recycling, )

Strategies for minimization of power requirements

DIII-Gohil-PRL01

JET-Andrew

PLH reduction by 20-30 % with pellets


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H-mode Hysteresis

Assessment of influence of local parameters versus power requirements and

role of ELM dynamics in H-L transition required

H-mode hysteresis results vary widely from experiment to experimentJET-Andrew-PPCF08

DIII-D-Thomas-PPCF98


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ITER operation in H-mode with edge power flux just above H-mode transition couldbe complex if JET-like behaviour reproduced in ITER

Cyclic transitions between Type I and Type III ELMy H-mode or even L-modeWplasmaoscillations > 20% P ~ Wplasma2 P oscillations > 40% amplification of Wplasma oscillations Problems sudden & large Wplasma excursions (possible large power fluxes to inner wall

due to radial plasma movement), control of divertor power flux under 10 MWm-2, additional

power coupling with oscillatory edge plasma conditions, etc.

JET-Sartori PPCF 2004 JET-Horton NF 1999

Lmode

Type I

Type

III

Open R&D : H ~ 1 sustainment in ITER (III)


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Start-up:q|| ~ 25 MWm-2, q|| ~ 5.0 cm

Several seconds

Confinement transients

q|| ~ 250 MWm-2, ~2-3 secs

Start-up and rampdown:q|| ~ 40 MWm

-2, q|| > 1.2 cm

Several seconds

VDE (up):q|| ~ 70-270 MJm

-2, q|| > 3.0 cm

t = 1.5-3.0 ms

VDE (down):

q|| ~ 90-300 MJm-2, q|| > 3.0 cm

Steady state:

q|| ~ 8 MWm-2

, q|| > 4.0 cmq|| ~ 24 MWm-2, q|| > 2.5 cm (ELMs)

Disruptionsq|| ~ 45-120 MJm

-2, q|| > 20 cm

t = 3.0-6.0 ms

Radiation:

SS: 0.5 MWm-2(photon+CX)

DisruptionsTQ: ~0.5 MJm-2

t ~ 1 ms (mitigated)

CQ: ~0.9 MJm-2

t ~ 10 ms

Distribution of FW panel design heat load


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Examples: major disruption and VDE on FW

Large areas receive energy densities > 10 MJm-2

Severe melting for either Be or W

13 MJm-2

Mitteau / Labidi

22 MJm-2

Full energy VDE

MD with

WTQ.=175

MJ

Pk factor =

3


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Examples: major disruption and VDE on FW Thermal specs. feed into lifetime estimates and requirements on mitigation

performance and success rate

Better guidelines also required on expected material losses13 MJm-2

Mitteau / Labidi

MD with

WTQ.=175

MJ

Pk factor =

3


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Disruptive load data is sparse and variable A few sparse datasets from a handful of devices

Great deal of heat load variation seen in different disruptions Strike point motion, splitting and non-axisymmetric at TQ MHD Captured only crudely by

broadening factor

JET, main chamber loads

Hollmann 12th ITPA DivSOL, San Diego

DIII-D

divertor

loads

Arnoux, NF 49 (2009)


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pre-TQ TQ C

Q

CQ

Radiation asymmetries during MGI

A. Huber, E. Hollmann PSI 2010A. Kallenbach, M. Reinke, 13th ITPA

ITER needs to estimate the extentof main wall heating by theradiation flash penalty if too

localised required no. ofinjectors

C-Mod

pre-TQ

pre-TQ

TQ CQ

JET10%D2

90% Ar

AUG

Ne

pre-TQ

pre-TQ

TQ CQ

DIII-D

Ne

Toroidal asymmetries


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Runaway electrons Heat load data extremely limited

Simple extrapolation to ITER from single JET discharge Must improve this situation

Lehnen, JNM 390-391 (2009)

Wetted area = 0.3 m2

in JET

0.3 0.6 m2

in ITER RE beam energy ~20 MJ 35-70 MJm-2 in ITER Need 6 - 14 MJm-2 to melt layer down to penetration depth in Be (2.5-7.5

mm for 1- 3 and 12 MeV)


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Secondary divertor ELM fluxes ELM filaments far from 2nd strike

20% ofWELM to 2nd strikeELM power even seen at inner 2nd strikeHigher than assumed in ITER load spec

DIII-D#

138219

Before

ELM

During

ELM

IR TV

DIII-DSecondar

ystrike

J. G. Watkins, IAEA 2010

More work required here all linked to understanding ELM broadening


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Prescribed ITER far-SOL inter-ELMprofiles critical for FW heat fluxestimates wall design

Assume break to convective(filamentary) transport in primary SOL

Based on tokamak data No predictive capability from current

models

Are ITER upper (high density)estimates correct?

What does the far-SOL look like with RMPs?

Open R&D : Far SOL heat fluxes

iter-scientific-status elms and disruptions

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