the efda programme duarte borba

2012 Annual Seminar of Association EURATOM-Tekes Estonia 29-30 May 2012

The EFDA Programme

Duarte Borba


OUTLINE

• Introduction to EFDA• EFDA Programme

– Latest Results and Future Programme • JET• ITER Physics• Power Plant Physics and Technology (PPPT)

• Summary and Conclusions


European Fusion Development Agreement (EFDA)

• Agreement between all EU fusion labs and Euratom

• Coordinated research (physics in support to ITER, longer term technology) and JET

• Garching (D) and Culham(UK)

Associations:European Fusion Laboratories associated to Euratom through “Contracts of Association”

The overall fusion programme is coordinated in the frame of EURATOM

Joint Undertaking for ITER “Fusion for Energy” (F4E)

• Domestic Agency to provide and manage EU contribution to ITER

• EU Contribution to Broader Approach

• Located in Spain (Barcelona)

Organization of EU programme


• Collective use of JET (~60M€/y operation + ~5M€/y exploitation + enhancements)

•Coordination of focussed physics and technology activities (~5M€/y priority support mobilizing ~25M€/y)

• Training (goal ~ 40 new trainee/y)

•EU contributions to international collaborations outside F4E

All EU Laboratories/Institutions working on Fusion are parties to EFDAEFDA

JET operated through a JET Operation Contract between the European Commission and the UK Atomic Energy Authority


EFDA Organization

EFDA Leader/Associate Leader for JET

F. Romanelli

Senior Advisor (Int. Collaborations) D. Borba

JETL. Horton

Public Information

P. Nieckchen

AdministrationC. Soltane

Power PlantPhysics &

Technology G. Federici

ITER Physics R. Neu

(From July 2012)


OUTLINE





Carbon era of JET ended in 2009

8


9

Be/W ITER-like Wall completed


10

Bulk W

W-coated CFC


Neutral Beam Enhancement Project

Beam voltage (kV)

Inje

cted

deu

trium

pow

er (M

W)

40 60 80 100 120 1400.0

0.5

1.0

1.5

2.0

2.5

Design goal: > 2.13 MW

EP1 PINIs

EP2 PINIs

PINI PerformanceNeutralisation efficiency measurements on Octant 8 beamline confirmed

that the design goal of >2.13 MW per PINI (>34 MW of total NBI power) can be achieved.

View through assembled duct in Assembly Hall, and installed in Torus


Long Term Fuel Retention Predicted for ITER

10x

(4 months)


Long Term Retention: rate normalised to divertor time

JET wall temperature

10x

JET-ILW: Is the absolute value low enough? True long term value could be much lower (surface analysis)

ITER

JET C-wall & ILW

10x

ICRHH-mode

Type III

NBIH-mode

Type I

L-mode


Scenarios: Hybrid plasma

N~2.810%

n/nGreenwald~80%

Zeff~1.2

Central Te

Line integrated ne

Total radiated power

H98

JET-ILW: H98~1 also achieved in low and high shape inductive scenarios

H98 ~1 requires low fuelling but W control restricts operating window

Next: Higher power, ELM pacing and central heating (as AUG-W)


LH Threshold with ILWJET-ILW: Total power threshold JET-ILW: Separatrix power threshold

ILW shown unpredicted strong effect on pedestal and ELM behaviour


2011 2012 2013 2014 2015 2016 2017 2018 2019 2020

Launched Proposed

ConstructionITER

Phase 1

Construction

JET

JT60SA

Possible Scenario (2011-2018/20)

Ph3 DT

Operation

SD SD *SD Ph2

SD = Shut down

Ph2SD

* Exact duration to be quantified

Ph3 DTITER preparation with tritium plasmas

Start of Enhancement Projects RMP

Under further discussion

Interleaved deuterium, full tritium, trace tritium and high neutron yield DT Campaigns

Commissioning In Hydrogen

Two possible enhancements have been studied:Electron Cyclotron Heating System (ECRH)Resonant Magnetic Perturbation (RMP) coils

ITER preparation with all the control tools foreseen in ITER

Commissioning and Joint experiments

ECRH


Plasma Stability and Control• Can large (~1MJ) ELMs be

suppressed ?

• Do n=4 and n=3 Resonant Magnetic Perturbations have the same effect ?

• Does a Be wall affect the ELM suppression ?

• Does the RMP affect the alpha particle confinement ?

Edge Perturbation Coils for Plasma Stability

Coils independently driven (in pairs)

at up to ± 60 kAt


Possible JET DT experiment after 2015

• Test ITER operation regimes with tritiumabout 10 years before first ITER DT experiment.

T

D

10 MW NBI

pede

stal

pre

ssur

e (k

Pa)

Type III

Type I

• Isotope dependence of confinement• ITER relevant Ion Cyclotron heating techniques with tritium• Diagnostic techniques relevant for DT operation in ITER


OUTLINE





, ,

VI-1: Diagnostics, 500

Reserve, 1600

VI-3: Fusion Materials Development, 600

III-2: Transport, 300III-1: MHD, 200

III-3: H&CD and Fuelling Physics, 130

V-1: Integrated Tokamak Modelling,

700

IV: Coordinated Activities on Plasma Wall Interaction, 750

VII-1: SERF, 20

Elec

tric

pot

entia

lan

d Io

n st

ream

line

s

shadowed

open

~1m

m

0.5

mm

Elec

tric

pot

entia

lan

d Io

n st

ream

line

s

shadowed

open

~1m

m

0.5

mm

EFDA ITER Physics Programme

• EFDA/F4E Collaboration on JT60SA EU Physics Activities • High Performance Computing


Modelling of RMP screening with plasma flows

First tests results on screening of RMPs by toroidal and diamagnetic flows were obtained for JET-like parameters (EFCC coils, n=2,96kAt),

Equilibrium radial electric field with diamagnetic 2 fluids effects in JOREK.

vacuum case

Islands screening by rotation

CEA

MHD


• Large Eddy Simulation tehcnique applied to gyrokinetics in GENE [Banon Navarro PRL 11, Morel PoP 11 & 12]

• Progress towards reliable multi-scale simulations at reduced numerical cost

ULB, IPP

[ Morel PoP 12 ]

Free energy injection spectra for the fourth-order model (M4) at reduced resolution, compared with a highly resolved DNS.

Multi-scale turbulence (modeling)Transport


Multi-scale turbulence (experiment)

• TORE SUPRA: measurements of extended wave number spectra enable validation of models for multi-scale turbulence, and turbulence simulations

GENE

CEA

[ Vermare PoP & EPS 2011 ]

Transport


Assumed SOL Density Profile

Modeling of scrape-off plasma with TOPICA and TOPLHA validation of the ITER and other IC and LH antenna design in plasma regimes with different scrape off layer density decay length. AUG antenna has been simulated in details.

Modelling of RF Sheath and power coupled to Plasma

ENEA

Modelled Power coupled to plasma

Heating and Current Drive


High-Z Materials Limiter with tungsten lamellae insert before exposure to TEXTOR plasma boundary

FZJ

Measured height profile along W surface after melting by excessive heat flux in TEXTOR discharge with computed height profile obtained by a MEMOS code simulation.

PWI Task Force


26

Anatomy of Modular codes

Legacy software structureITM Task Force


38th EPS 2011, June 27 – July 1, C. Konz

27

Standardized module interfaces:

• allow for simple construction of generic workflows

Access to database

Equilibrium chain Kepler Workflow



28



• easy interchange of modules from same type of physics

Equilibrium reconstruction: free boundary equilibrium module




29



• easy interchange of modules from same type of physics

Equilibrium refinement: high resolution fixed boundary equilibrium module




30

Generic visualization using Python scripts




• allow for generic visualization of results, e.g. Python, VisIt


Demonstrated Kepler workflow, using experimental/synthetic JET data, coupling the equilibrium code Helena and the gyrofluid turbulence code GEM. GEM is executed in batch on the HPC-FF

First-principle turbulence codesworkflow running on HPC


Core-Edge Coupling

• Te in the Scrape-Off Layer computed by SOLPS

• Te in the core computed by ETS (coreprof edge)

• Visualized in the ASDEX Upgrade device geometry

• Raw geometry data provided by IPP Garching (T. Lunt),

• Postprocessing done at Aalto University, Helsinki (T. Kuoskela)

• Simulation and integrated visualization performed at IPP Garching (H.-J. Klingshirn),

• Using VisIt visualization software and general-purpose ITM tools


Reflectometer plasma position controlReflectometer plasma position control was achieved both in L and type-I ELMy H–Mode phases.

The shaded regions show the discharge phases in which reflectometry separatrix estimate replaced the magnetics as the radial outer plasma position source for feedback control.

The solid black lines show the pre-programmed plasma position controller target trajectory. The green curves show the magnetic separatrix position and the red the reflectometry estimates for the outer separatrix position.

IPP, IST

Diagnostics and RT Control


EFDA/F4E Collaboration on JT60SA EU Physics Activities

• EU Research Unit Coordinator is Gerardo Giruzzi (CEA).• Joint EU-JP revision of the JT60-SA research plan

completed in 2011 with the issue of version 3.0. • Meeting to finalise the activities for 2012 will take place

in in Frascati, 4 – 5 June 2012.


EFDA/F4E Collaboration on JT60SA EU Physics Activities

2012 Work Programme focus on: -General subjects

- Follow-up of Research Plan revision,- Modelling of JT-60SA plasmas (performed by the ISM project ) - Definition of JT-60SA data system, validation and analysis tools- Experiments and data analyses on EU machines in support of JT-60SA

- Specific subjects - Evaluation/optimization of the ECRH launcher - Assessment of selected diagnostic systems - Assessment of JT-60SA divertor pumping system


High Performance Computing HPC-FF Julich, Germany

Comparison of the filamentation of the plasma during an ELM in the MAST tokamak

• Total of 68 projects awarded CPU time in 2011 and 60 projects in 2012


Allocated CPU-hours

Turbulence

Materials

Edge

MHD

Nuclear

RF

Mix

Fast particles

High Performance Computing HPC-FF Julich, Germany

HPC-FF Usage since Production Started in August 2009

•HPC-FF now basically full (some idle time necessary to queue-in large CPU jobs)


High Performance Computing IFERC Computer (Helios), Rokkasho, Japan

• According the Terms of Reference of the Standing Committee governing the usage of the IFERC supercomputer in Rokkasho, 20% of the time will be allocated directly to the parties (10% for each of the EU and Japan) to enable them to do tests, tuning or adaptation of codes, etc. The parties will internally organise the use of this time, which is available since 9th April 2012.

• The remaining time 80% is allocated to projects.

• 58 Proposals were accepted and allocated 20.45 M Node Hours in the 1st Cycle of operation

• Deadline for proposals for the Second Cycle is end of July 2012.

http://www.iferc.org/csc/for-researchers/hpc-information.html


High Performance Computing IFERC Computer (Helios), Rokkasho, Japan

• After discussions with F4E it is agreed that the 10% of time for the EU will be allocated according such as each Associate receives

800 000 CPU core hours (50 000 node Hours), for use by fusion researchers in their country.

• The same contact persons, as used for the HPC-FF Association accounts, administer the Helios accounts.

For TEKES the contact person is Jukka Heikkinen ([email protected])


OUTLINE





System code Studies for optimisation of System code Studies for optimisation of Fusion power plant DEMO conceptsFusion power plant DEMO concepts

• For pulsed operation, the main aspects of concern at present are the flux swing, pulse length and divertor heat load modelling, although there are some concerns about cyclic fatigue issues, which are not well captured in systems studies at present.

• In modelling steady-state devices, the main concern at present is the divertor heat load and radiation modelling.

• These studies should lead to improved modelling but also to new experimental work to validate key assumptions such as the impact of radiation on confinement, plasma behaviour at high density and high β.

Conservative Advanced


Assess Prospects for Alternative Assess Prospects for Alternative Concepts for Heat Removal in Fusion ReactorsConcepts for Heat Removal in Fusion Reactors

• Capillary porous systems (CPS)– Optimize heat extraction without evaporation cooling for Li systems in

magnetic confinement devices. – Test the resilience to transients loads (disruptions, ELMs)– Test other material options (e.g. Ga, Al, Sn)

• Droplet systems– Check stability of droplet systems in magnetic confiment devices and in

test stands with high magnetic fields

• Liquid metals, basic properties– Identify temperature window with low evaporation and low tritium retention

(mainly Li)– Understand T retention mechanism– Decide on reactor compatibility


High Temperature Super ConductorsHigh Temperature Super Conductors• REBCO (rare-earth barium-copper-oxide high-temperature

superconductors) is the only candidate for application for fusion magnets

• Demonstrators have shown currents up to 7.5 kA (77 K, self field)

• Operation at 4 K and fields in the range of 12 T - 18 T is possible and allows

• e.g. 74 kA @ 4 K and 15 Tesla • 30% volume reduction of winding pack• Effective current closer to the plasma -> higher field

• The higher cost for REBCO could be partially balanced by simpler structure (no radial plates), no SC annealing andsimpler HV insulation of the cable


Remote HandlingRemote Handling• In-vessel gamma radiation dose levels of ~20kGy/hr after 2 fpy will

prohibit the operation of many remote handling technologies within the Tokamak

• The Vertical Maintenance System with Multi Module Segment (MMS) is considered to be the most viable plant architecture for power plant relevant maintenance

• Due to the harsh conditions predicted within the tokamak, ex-vessel remote operations need to be maximised to minimise those required in-vessel and planned in-vessel maintenance should not include any complex operations, i.e. cutting and welding.


Material Programme in EFDAMaterial Programme in EFDA

The materials programme in EFDA is being reoriented along three main lines:(a)DEMO materials choices and specifications:

• DEMO 2030 will probably need to use “existing” materials (e.g. EUROFER)• Reconsider Cu-alloy for a water-cooled divertor.

(b)guidance of test matrixes for irradiation campaigns• design of material irradiation experiments to consolidate the materials

database for DEMO structural materials using existing fission reactors (e.g., by using isotopic tailoring) dual ion-beams, modelling of radiation effects, exploitation of post-EVEDA irradiation campaigns

(c)development of improved materials for later fusion devices• Tungsten and tungsten alloys• Oxide-dispersion-strengthened ferritic steels (ODSFS)• Material modelling.


SummaryJET • The ILW JET shutdown was completed • First results confirm low tritium retention• Operation with the ILW achieved more easily then expected• Good H-mode confinement discharges established

The ITER-like Wall has shownAnticipated benefits and risks of a W/Be wall:Large reduction in fuel retention and very clean/reproducible plasmasITER scenarios constrained by W-accumulation but still achievableGood power handling and protection of the Be/W wall

An unpredicted strong effect on pedestal and ELM behaviour


Summary• ITER Physics EFDA programme covers a number of key physics priority questions for ITER exploitation:

• MHD, Transport, H&CD, PWI, ITM, Diagnostics

• Preparation of the operation of JT60-SA together with Japan• Management of Super Computer Resources (HCP-FF and Helios)

• Power Plant Physics and Technology (PPPT) activities:• System code Studies for optimisation of Fusion power plant DEMO concepts• Concepts for Heat Removal in Fusion Reactors • High Temperature Super Conductors for Fusion Applications• Remote Handling• Fusion Materials Research


Thank you for attention


Backup Slides


EFDA 2012 Work Programme Implemented under Cross Topical Areas of Research

1. Prediction of Material Migration and Mixed Material Formation (PWI, DIA)

2. Shaping and controlling performance limiting instabilities (MHD, DIA, H&CD-F)

3. Fuel Retention and Removal (PWI, DIA, H&CD-F) 4. Plasma rotation (MHD, TRA, H&CD-F)5. Core electron heat transport and multi-scale physics (TRA,DIA)6. Pedestal instabilities (ELMs), Mitigation and Heat loads (MHD,

TRA, PWI, DIA, H&CD-F) 7. Disruptions, prediction, avoidance, mitigation and consequences

(MHD, PWI, DIA) 8. Physics of the Pedestal and H-mode (DIA, TRA, H&CD-F) 9. Fast Particles (MHD, TRA, DIA, H&CD-F)10. Particle transport, fuelling and Inner Fuel Cycle modelling (TRA,

H&CD-F, PWI) 11. Operation with metallic plasma facing components, including High

Power ICRH (PWI, H&CD-F)


New CC Chairs for H&CD• Two new Chairs of H&CD Coordinating Committees have been appointed:

– J.M. Notraedame (IPP) for the Coordinating Committee Ion Cyclotron Heating, replacing R. Koch;

– A. Tuccillo (ENEA) for the Coordinating Committee Lower Hybrid, replacing T. Hoang

– and E. Surrey (CCFE) chair of the Coordinating Committee on Neutral Beam– C. Day (KIT) chair of the Coordinating Committee Fuel and Pumping


Calibrated particle source(gas injection modules)

Gas collection on divertor cryo pumps

Fuel retention: long (co-deposition & implantation) +

short term (surface)

Repeat sets identical discharges

Injected gas = Pumped gas+ Short term retention + Long term retention

Regeneration of cryogenic pumps before and after experiment

1. Stronger gas consumption during limiter phase with ILW ≈ 2x2. Lower gas consumption during divertor phase w.r.t. CFC walls3. Stronger outgasing after the discharge compared to CFC walls

Note: C-wall long term retention from surface analysis << gas balance

Fuel retention: JET Gas Balance Experiments


High Frequency Pellet InjectionSmall pellets 1.5mm diameter

Large pellets (4mm diameter)

Highest velocities not achieved for highest injection frequencies (especially for large pellets)


Commissioning with Plasma 2012• Ice quality ‘oscillates’ during extrusion. • Good pellets only during periods of good ice• Hard to synchronise pellets with plasma.

Nevertheless: Some successful experiments with large pellets:

Achieved in one shot increased ELM frequency 7 Hz to 20 Hz at averaged, up to 31 at maximum (short phase): Unwanted fuelling side effect gone for entire sequence!


55

9. ConclusionsThe ITER-like Wall has shownAnticipated benefits and risks of a W/Be wall:Large reduction in fuel retention and very clean/reproducible plasmasITER scenarios constrained by W-accumulation but still achievableGood power handling and protection of the Be/W wall

An unpredicted strong effect on pedestal and ELM behaviour

ITER/DEMO scenario development needs relevant PWIi.e. PWI issues and H-mode physics are not separable

Exploitation & analysis of the ITER-like Wall has only just begun


Fuel retention

I II III

2.3μm

TEM image of cut through surface zone of W sample irradiated by 20 MeV W+ with damage sites visible to a depth of up to 2.3 mm (indicated by red arrow). Three zones can be separated: I – near surface area with high density of dislocations, II – intermediate area with lower density of “long” dislocations and “sub-grain” regions free from dislocations, III – deep region with high density of “short” dislocations.

• Apart from material structure, fuel retention in tungsten is also strongly influenced by radiation damage, increasing the number of trapping sites for fuel atoms in the material. • The type of damage sites created by this method was studied by cutting lamellas out of W-ion irradiated tungsten samples and TEM analysis of the lamellas.

IPP


Fuel retention

ED+=20eV

0 1 2 3 4 510-3

10-2

10-1

100

D @ vacancies /vacancy clusters

D @ dislocations

damageprofile 630 K

500 K

damaged W exposed to 2.2×1025 D/m2

Dco

ncen

tratio

n /a

t.%

Depth / μm

IPP

Depth profiles of implanted deuterium in pre-damaged tungsten measured by nuclear reaction analysis for different sample temperatures during low-energy D-plasma exposure. With increasing temperature, D does not decorate dislocation sites anymore but is still trapped at vacancies created by previous radiation damage within the first 2 microns of the surface.


Transient heat loads CCFE

Fraction of stored plasma energy reaching divertor surface after mitigated disruptions D using Neon, Argon or He/Ar mix in MAST.

Compared to typical deposition of 80% of stored energy in the divertor following unmitigated disruptions, massive gas injection can reduce this fraction down to 40%.


• A dual probe technique comprising planar Langmuir probe and hairpin probe has been tested and calibrated with Laser photo detachment technique to estimate the negative ion density.

• The negative ion density in front of anodic glow is on the orders of 1011 cm-3 and increases monotonically with applied power.

NBI source development (negative ions plasma source)

Plasma Source

Spatial plasma parameters are measured from the edge of the anode glow, using a hair pin probe and emissive probe.

IPPHeating and Current Drive


• Heat loads onto divertor of 10-15 MW/m2 require a heat sink material with high thermal conductivity: Only Cu-alloys can be considered

• In 2011 an initial assessment of data available on Cu-alloys was conducted (PEX/02 + DAS MAT).

• Lack of data for relevant irradiation doses. • Need to investigate also high-radiation resistance Cu-alloys

Water-cooled Divertor

the efda programme duarte borba

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