the w7-x team - irfmirfm.cea.fr/isfrt16/lectures/20170503_dinklage_erice_rev.pdf · • (some...

Alonso[1], Andreeva[2], Baldzuhn[2], Beurskens[2], Beidler[2], Biedermann[2], Blackwell[17], Blanco[1], Bosch[2],

Bozhenkov[2], Brakel[2], Burhenn[2], Buttenschön[2], Cappa[1], Czarnetzka[3], Dinklage[2], Endler[2], Estrada[1],

Fornal[3], Fuchert[2], Geiger[2], Grulke[2], Hartmann[2], Harris[4], Hirsch[2], Hoefel[2], Jakubowski[2], Klinger[2], Klose[2],

Knauer[2], Kocsis[5], König[2], Kornejew[2], Krämer-Flecken[6], Krawczyk[3], Krychowiak[2], Kubkowska[3], Kiazek[7],

Langenberg[2], Laqua[2], Laqua[2], Lazerson[8], Maaßberg[2], Marsen[2], Marushchenko[2], Moncada[9,10], Moseev[2],

Naujoks[2], Otte[2], Pablant[8], Pasch[2], Pisano[11], Rahbarnia[2], Riße[2], Rummel[2], Schmitz[12], Schröder[2],

Stange[2], Stephey[12], Szepesi[5], Trimino-Mora[2], Thomsen[2], Traverso[13], Tsuchiya[14], Turkin[2], Velasco[1],

Wauters[15], Werner[2], Wolf[2], Wurden[16], Zhang[2], et al.

[1] CIEMAT, Madrid (Spain) [2] Max-Planck-Institut für Plasmaphysik, Garching, Greifswald (Germany) [3] IPPLM, Warsaw (Poland) [4] Oak-Ridge National Laboratory, Oak Ridge, TN (USA) [5] Wigner RCP, Budapest, (Hungary) [6] Forschungszentrum Jülich, Jülich (Germany) [7] Opole Univerisity, Opole (Poland) [8] Princeton Plasma Physics Laboratory, Princeton, NJ (USA) [9] CEA, Cadarache (France) [10] ThermaVIP Ltd., Cadarache (France) [11] University of Cagliari, Cagliari (Italy) [12] University of Wisconsin, Madison, WI (USA) [13] Auburn University, Auburn, AL (USA) [14] National Institute for Fusion Science, Toki (Japan) [15] ERM, Brussels (Belgium) [16] Los Alamos National Laboratory, Los Alamos, NM (USA) [17] Australian National University, Canberra (Australia)

Andreas DINKLAGE for the W7-X Team | First Experiments on W7-X | NIFS, Toki | 30. May 2016 |

The W7-X Team

T

Theorist‘s view on Wendelstein 7-X

(from its top; blue: coils, yellow: plasma surface)

Summary

• Optimized Stellarators: a potential path to a fusion power plant (FPP )

• W7-X is to bring the HELIAS line to maturity (FPP) Qualify key technology

Show good plasma confinement & proof optimization

Demonstrate safe high-performance steady-state operation

(heating, fuelling, exhaust, impurities, fast-ions, MHD, scenarios)

• (some selected) diagnostics, their requirements and applications: results from the first operation phase Diagnostics are very much as in tokamaks,

accents on 3D, steady-state, ECRH possible contributions to ITER, DEMO

W7-X is comprehensively equipped with diagnostics allowing to address first aspects of the high-level vision of W7-X

• Diagnostics for S-DEMO?

Andreas DINKLAGE | Int. School of Fusion Reactor Technology | ERICE, Italy | 03. May. 2017 |

This work has been carried out within the framework of the

EUROfusion Consortium and has received funding from the

European Union‘s Horizon 2020 research and innovation

programme under grant agreement number 633053.

The views and opinions expressed herein do not

necessarily reflect those of the European Commission.

Diagnostics Developments for W7-X (not all!)

Potential Applications for Burning Plasma Devices

International School of Fusion Reactor Technology, 03.05.2017

Andreas Dinklage1,2 for the W7-X Team 1Max-Planck-Institut für Plasmaphysik, Greifswald, Germany 2E.-M.-Arndt Universität Greifswald

with gratitude to R. König, D. Hartmann and S. Bozhenkov

Alonso[1], Andreeva[2], Baldzuhn[2], Beurskens[2], Beidler[2], Biedermann[2], Blackwell[17], Blanco[1], Bosch[2],

Bozhenkov[2], Brakel[2], Burhenn[2], Buttenschön[2], Cappa[1], Czarnetzka[3], Dinklage[2], Endler[2], Estrada[1],

Fornal[3], Fuchert[2], Geiger[2], Grulke[2], Hartmann[2], Harris[4], Hirsch[2], Hoefel[2], Jakubowski[2], Klinger[2], Klose[2],

Knauer[2], Kocsis[5], König[2], Kornejew[2], Krämer-Flecken[6], Krawczyk[3], Krychowiak[2], Kubkowska[3], Kiazek[7],

Langenberg[2], Laqua[2], Laqua[2], Lazerson[8], Maaßberg[2], Marsen[2], Marushchenko[2], Moncada[9,10], Moseev[2],

Naujoks[2], Otte[2], Pablant[8], Pasch[2], Pisano[11], Rahbarnia[2], Riße[2], Rummel[2], Schmitz[12], Schröder[2],

Stange[2], Stephey[12], Szepesi[5], Trimino-Mora[2], Thomsen[2], Traverso[13], Tsuchiya[14], Turkin[2], Velasco[1],

Wauters[15], Werner[2], Wolf[2], Wurden[16], Zhang[2], et al.

[1] CIEMAT, Madrid (Spain) [2] Max-Planck-Institut für Plasmaphysik, Garching, Greifswald (Germany) [3] IPPLM, Warsaw (Poland) [4] Oak-Ridge National Laboratory, Oak Ridge, TN (USA) [5] Wigner RCP, Budapest, (Hungary) [6] Forschungszentrum Jülich, Jülich (Germany) [7] Opole Univerisity, Opole (Poland) [8] Princeton Plasma Physics Laboratory, Princeton, NJ (USA) [9] CEA, Cadarache (France) [10] ThermaVIP Ltd., Cadarache (France) [11] University of Cagliari, Cagliari (Italy) [12] University of Wisconsin, Madison, WI (USA) [13] Auburn University, Auburn, AL (USA) [14] National Institute for Fusion Science, Toki (Japan) [15] ERM, Brussels (Belgium) [16] Los Alamos National Laboratory, Los Alamos, NM (USA) [17] Australian National University, Canberra (Australia)

The W7-X Team


1. Introduction: stellarators in a nutshell

2. Measuring W7-X

Results from the first operation phase of W7-X

Diagnostics developments and aspects relevant to steady-state

operation

3. Summary

contributions to BPXs and open issues for stellarator reactors

Outline


• What is a stellarator and how does it work?

• What is the potential of stellarators for a future fusion

power plant?


I. Introduction

Magnetic Confinement

How to get to equilibria for toroidal plasmas?

Need for rotational transform and flux surfaces

What is a stellarator?


Tokamak vs Stellarator?

Stellarator

Two main approaches to twist

toroidal magnetic fields: ‘gas engine‘ and ‘diesel’ of fusion.

Tokamak Tamm Sacharow Spitzer

Stellarators: + steady-state, no large plasma currents,

- 3D losses/engineering, one generation behind



Tokamak

Confinement & symmetry

large toroidal currents generate rotational transform

E. Noether (1882-1935)

If a system has a continuous

symmetry property, then there are

corresponding quantities whose

values are conserved in time.



0

Stellarators

3D confinement: helical fields modulate B toroidally

Beidler et al Nucl. Fusion 51, 076001 (2011) (2001)

BxB and

curvature drifts:

locally trapped particles

are quickly lost!

q

p

0

polo

idal angle

2p

toroidal angle

p/N 2p/N



Why Wendelstein7-X?

The HELIAS* concept and its scientific perspective

Assess stellarator optimization: overcome classical stellarator‘s draw-backs

Bring stellarators to maturity: understand hot plasmas in 3D

© Kleiber, Borchardt et al.

*Nührenberg, Zille, Phys. Lett. A 114, 129 (1986)

How to get from the idea to a power plant?



Is there a reasonable perspective for the stellarator line for

fusion energy?


„[T]here are known knowns; there are things we know we know. We also know

there are known unknowns; that is to say we know there are some things we

do not know. But there are also unknown unknowns – there are things we do

not know we don't know.“

Rumsfeld (2002)


© Schauer, Bykov, et al.

Known unknowns

Physics Optimization criteria (W7-X)

Divertor loading (W7-X)

Confinement scaling (W7-X)

Steady-state operation (W7-X)

Tritium breeding (Tokamak line)

Engineering Coil design (S-DEMO Engineering studies)

Blanket design (S-DEMO Engineering studies)

Vessel design (S-DEMO Engineering studies)

Support design (S-DEMO Engineering studies)

Blanket maintenance (S-DEMO Engineering studies)

Material issues (Tokamak line)

Economics Size optimization (S-DEMO Engineering studies)

Operational availability (S-DEMO Engineering studies)

Safety Tritium inventory (Tokamak line)

Decay heat (Tokamak line)


… but we have W7-X (and other stellarators), ideas & concepts

How far away is a reactor? How to build a stellarator reactor?

Tokamak Stellarator

? ISS, Q=10

Step ladder to fusion power plant

K. Lackner, Fusion Sci. Technol. 54, 989 (2008)


Coil design ITER and HSR5 coils (same scale)

• F Schauer et al., Contrib. Plasma Phys. 50 (2010) 750

ITER toroidal field (TF) coil HSR50a coil #5

ITER (TF only)

HELIAS 5B

Magn.field at plasma axis

5.3 T 5.6 T

Maximum magn. field

11.8 T 12.3 T

Superconductor

Nb3Sn Nb3Sn

Circumference

34.5 m 34.7 m

Minimum bend radius

2.0 m 1.63 m

Magn. energy per coil

2.3 GJ 3.2 GJ


Coil design for reactor scale devices w/ ITER technology

Vessel & support design for HSR

Design studies (Schauer et al, 2012)

No central support ring as in W7-X

The inter-coil structure

consists of bolted panels

One or two panels with one or

two plates depending on load

distribution

Panel size ~ 1 × 1.5 m2

Stresses within allowable limits for

stainless steel (1.4429) at 4K

Vessel: Double hull structure

similar to ITER VV

Wall thickness: >60 mm each

In-between water & steel

shield: 220 mm

Vessel

Support

Stresses


Blankets & Maintenance Concepts Blanket & Shield

• Size of HELIAS 5-B is determined

by blanket space requirements

• Space between coil and plasma:

1.3 m, blanket thickness: 80 cm

• 400 blanket segments

• Geometry not yet optimized

• Using half module symmetry

aiming at 40 different segments

• Flux surfaces are preserved

Maintenance

• One large vertical port per module

(5 modules) 4.3x(2.5, 1.8) m²

• Separation of plasma and outer

vessel

• Remote handling device for

separation (KIT)

• Insertion of maintenance boxes

Blanket

and shield

Stresses

Idea for

maintenance

approach


Where do stellarators stand?

• Stellarators – 3D magnetic confinement facility B generation by coils outside the plasma

+ steady-state, stable configuration w/o current, reversed shear

+ no disruptions; radiation collapse slower that tokamak disruptions

+ high-density operation possible lower pa at given Q

- 3D engineering – integration and maintenance

- concept development one generation behind the tokamak – many unknowns

• About stellarator DEMO control + less effort needed for real time control of current and plasma position

+ milder instabilities

- plasma scenario to be explored (confinement, impurity control)

- burning plasma effects unknown

- divertor operation and detachment control to be explored

Adapted from W. Biel et al., SOFT San Sebastian


W7-X is the key device for the stellarator line.

How does it specifically contribute to research and

operation of larger (burning) stellarators and what are the

potentials to contribute to developments for burning plasma

devices in general?


II. Measuring W7-X

W7-X: HELIAS en route to a FPP

Wendelstein 7-X

Basic research: ionized matter under extreme conditions

Energy research: confinement of hot plasmas for fusion

Mission: bring stellarators to maturity

Wendelstein 7-X

HELIAS-type stellarator

• Nf=5, R/a = 5.5m/0.53m

→ 30 m3 plasma volume

•heating

• ~8+7MW (ECRH, NBI)

• ICRH (~1.2 MW, later upgrades)

no DT operation:

plasma physics experiment

Greifswald (Germany) operating since 2015

-


• 70 superconducting coils (2.5T)

• 5 x 2 x 5 non-planar coils

• 5 x 2 x 2 planar coils

Wendelstein 7-X


Optimization and phys. requrmts. What is a stellarator?



Project time-line

Summary?

2015

OP 1.1

(13 Wks)

P < 5 MW

Te , Ti 3 keV, 1 keV

n 0.2 x 1020 m-3

< 1.6 %

Limiter configuration

Pulse limit

P dt 2 MJ

pulse ~ 1 s

T.S. Pedersen et al., Nucl Fusion 2015

First experiments on Wendelstein 7-X

Goal of the first operation phase[1]:

1. demonstrate the existence of flux-surfaces

2. integral commissioning of a complex machine:

magnets, cryostat, heating, diagnostics, control & data acquisition

… and if there was time to do some experiments …

Initial plan: demonstrate operation and do first

measurements


Andreas DINKLAGE for the W7-X Team | Wendelstein 7-X | Universität Siegen | 16. Jun. 2016 |

Milestone #1: proof of flux surfaces

© M. Otte

Can W7-X (a HELIAS) be built?

B11 is within the allowed margins so far.

TS. Pedersen et al. Nature Comm. (2016)

W7-X Camera Systems


Milestone #2: First plasma

Dec. 10th, 2015

First helium plasma in W7-X was created according plan

Full field: B = 2.52 T

PECRH = 1.3 MW

pulse = 50 ms

First measurements of

plasma parameters conducted

Te ~ 100eV

ne ~ O(few 1019 m-3)

First shot program (sequence of conditioning pulses)

w/ steady-state control system conducted on Dec. 11th

Can a HELIAS be operated? Yes.



Safe Operation: Camera Systems


W7-X Camera Systems


Milestone #3:First Hydrogen Plasma in W7-X

First hydrogen plasma in Wendelstein 7-X

total heating power: ~ 2 MW

pulse lengths: ~ 250 ms

Te ~ 7 keV, Ti ~1.2 keV

<n> ~ 2 x1019m-3

a < 49 cm (V ~ 26 m3) (plasma touched the limiter)

3.2.2016, 15:21:25.822 (local time) … 60 s later

W7-X Team | Greifswald, Germany | 03.Feb.2016


time (s)

limiter current (a.u.)

Ha(neutrals, a.u.)

Tion (keV)

Te (keV)

P (MW)

absorp.

ne dL (1019m-2)

2

0

10

0

0

1

2

0

0 0.3

• limiters not overheated even in 2 MJ discharges,

4 MJ per discharge was allowed during the last weeks of operation

• From 1 s to 6 second discharge shown (1 s 1MW, then 5 s 0.6 MW):

Prolongation of discharge duration First experiments on Wendelstein 7-X

time (s) 0 1 2 3 4 5 6

line density

ion temperature

electron temperature

heating power

ne d

L

(10

19m

-2)

Ti

(keV

) T

e

(keV

) P

ech

(MW

)

2

0 10

5

0

4

2

0

0

1

2

1


Overall performance


Confinement for first plasmas not degraded w.r.t. tokamaks


Diagnostics for plasma core studies

* to come


Magnetics


Magnetics for long pulses

3D aspects

Folded loop made of continuous ribbon cable fits through large ports for installation

In-vessel assembly test without ECRH stray radiation shield

Foldable Diamagnetic Loop to Avoid Thermo-voltages at Connection Plugs

Steady-state electronic drift compensation

drift & common mode signals eliminated by chopping and numerical processing

drift: 4 A/100s

Drift measured with 100 A wire

through Rogowski coil installed

on W7-X plasma vessel:

Lowest expected plasma

current 5 kA in W7-X

Excellent performance of integrator, well suited for 30 min discharges and low

signals in W7-X, max. drift 72 A over 30 min

A. Werner

Long pulse integration, all currents (PS, bootstrap) are small (W7-X Optimisation)



∫ nedl


∫ nedl for steady-state operation



@ TEXTOR

Dispersion interferometer module



CuCrZr

Mo

Mo

10 mm 5 mm

principle:

heatload to mirror 90 kW/m2

device fixed at bottom T_interface=270oC -> T_max=290oC -> smooth temperature distribution and deformation

deformation

T-distribution

293. o C

270. o C Dl = 50 mm

expected analytically for slab: Dl = 40 mm

fastened

heatload

cooling

simplified ANSYS model:

deformation across surface

0.03 mm / division

Dz = 0.150 mm

μm17.0Dz

analytical slab model delivered

expected: Dl = 2.9 cm * 5.1 10-6 *(282-20)oC = 40 mm

2 cm


Thomson scattering on W7-X


X-ray Imagings Spectrocopy Ti, Er


Ti

Heat loads

0 +100 +200 +400 +800 +1000 A

Heat load shifts upwards as the n=1 perturbation trim coil currents with a

maximum in Module 3, are increased (while holding the phasing fixed).

(c) G. Wurden, LANL


M.Jakubowski, S. Lazerson, et al.



Surveillance for safe operation

Thermal loads from radiation

Heat load distribution across the first wall at Prad = 10 MW

● 107 test photons

on flux surfaces

● 300.000 surface

primitives

(ANSYS mesh)

● 30 min. on

Linux PC

Protective measures: Water cooled SS heat shields, water cooled windows & mirrors

3D Monte Carlo simulations


S-DEMO control requirements table

Quantity Diagnostics Actuators Interactions Control

accuracy

Spatial

resolution

Control

time response

Main plasma density

Polarimetry Reflectometry Spectroscopy Neutrons

Gas injection Pellet injection Pumping system

Wall and divertor, temperatures (outgassing)

2 % - 5% (10%)

a/10 in core a/20 in edge

0.1 … 1 s for 10% in-crease, 1 … 5 s for 10% de-crease

Main plasma temperature

ECE Spectroscopy Neutrons

Aux. heating Gas injection

Main plasma density

5% - 10%

a/10 in core a/20 in edge

Several s for increase, A few ms for decrease

Plasma position and shape

Reflectometry ECE Magnetics behind blanket

PF coils CS coils Plasma heating

Confinement (beta)

a/50 a/100 > 0.1 s (PF coils) < 0.1 s (confinement)

Zeff, impurity composition

Spectroscopy Uloop

Impurity gas inlet

FW and Div fluxes, erosion

0.2 - 0.5

Integral or a/5 1 s

Fusion power FW and Div coolant temperature

Gas injection Pellet injection Impurity inject. Aux. heating

Confinement (beta)

Pmax/50 Integral Several s for increase, A few ms for decrease

Plasma instabilities

Reflectometry ECE

ECRH

q profile beta density Zeff

t.d.b. a/40 (t.b.d.) < 1 ms

Divertor detachment and heat flux control

Reflectometry ECE Spectroscopy Divertor current

PF coils Gas injection Pumping system

Confinement (beta)

t.b.d. t.b.d. 10 ms

Adapted from W. Biel et al., SOFT San Sebastian

?

∫ndl, ~n

nus

?

particle ~ few s

?

?

T0, Tus

?

?

?

?

skin,L/R ~ 10s ..

min

(feed-forward?)

?

? ?

• W7-X first phase: all technical and scientific objectives successfully achieved [1]

safe routine operation of ECRH, cryo-plant, coil system and control/DAQ

demonstrated

about 20 diagnostics successfuly commissioned and delivered results

allowed the W7-X Team to safely increase technical limits (2 → 4 MJ)

allowed one to change magnetic configurations

• opened the door for an unexpectedly comprehensive physics program

• First experiments on Wendelstein 7-X :

fundamental research: self-organization high electron temperatures, turbulence

assessing a potential path to a fusion power plant:

assembly accurate, flux surface, first step to steady-state operation

… and even more physics topics in view of future operation addressed:

confinement/transport, heating and drive of plasma current, tools for exhaust

[1]Sunn Pedersen et al., Nucl. Fusion 55 (2015) 126001

Summary I: W7-X Results

Andreas DINKLAGE for the W7-X Team | Wendelstein 7-X | Universität Siegen | 16. Jun. 2016 |

• Optimized Stellarators: a potential path to a FPP

• W7-X is to bring the HELIAS line to maturity OP1.1: good flux surfaces, neoclassical effects

• Physics and technological boundary conditions drive diagnostics Qualify key technologies

(modular coils, steady-state operartion, ECRH)

Show good plasma confinement & proof optimization

Demonstrate high-performance steady-state operation

(heating, fuelling, exhaust, impurities, fast-ions, MHD)

• Diagnostics for S-DEMO?


Summary II: Potentials

the w7-x team - irfmirfm.cea.fr/isfrt16/lectures/20170503_dinklage_erice_rev.pdf · • (some...

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