Conductors and Coils for nuclear fusion magnets: from state-of … · 2019-12-17 · L. Muzzi - Seminario Roma TRE - 5 Dic. 2019 17 NbTi Nb 3Sn B peak determines the choice of the

L. Muzzi / FSN – Superconductivity Laboratory

Conductors and Coils for nuclear fusion

magnets: from state-of-art LTS

technologies to perspective for HTS

Seminario Univ. Roma TRE

5 Dicembre 2019

Outline

2

!  Introduction on the Magnet system of a tokamak reactor

!  Superconducting strands and Cable-in-Conduit

conductors (CICCs)

! Manufacturing aspects of ITER CICCs and coils

! What’s beyond ITER?

!  DEMO

!  DTT (Divertor Tokamak Test Facility)

!  HTS-based fusion

L. Muzzi - Seminario Roma TRE - 5 Dic. 2019

Fusion magnets: the ITER coils

3 L. Muzzi - Seminario Roma TRE - 5 Dic. 2019

Magnetic confinement fusion: the tokamak concept




TF (Torodial Field): 12T- 68kA – steady state

CS (Central Solenoid): 13T- 46kA – transient

PF (Poloidal Field): 6.4T- 52kA – transient




ASG, Italy




General Atomics, USA




ASIPP, China

Outline




conductors (CICCs)



!  DEMO


! HTS-based fusion

Superconducting wires (strands): stability


!"#$%"&'$%()*)$$+,-%".%*.'/0)%

1!"#$%&%'(2%"&)%3+"),4+0%5.,%)6+3/0)%

74"&%)*++(,-.,%.&/0%1/0-

Superconductor Cu

α = Cu/nonCu ratio

In the normal state the superconductor, has high electrical

resistivity and low thermal conductivity

2kA(Tc-Top)/l = Jc2ρ!"#

Superconducting wires (strands): stability


!"#$%&'()(&*&#+,&&(&#-#.+/0#1/23*45!

P =nτ

µ0

dB

dt

2

τ =µ0

2ρet

p

2π

2

!"#6,/3+*45#+,&&(&!

Multi-filamentary superconducting wires


!"#$%&'()"&*'%+,'##$%+

-.$+/"#$%&'()"&*0+12%$+2/+

3'%4$)+'3+45(6+!"#$%&'()*$!++

'3+/0&0+45*$%257/8+12*.2(+5+,-%

/*59272:2(;+)(!.#/%

Wire diameter 0.5÷1 mm

# supecond. filaments 1000÷10000

Filament diameter 5÷50 µm

Cu/non Cu 4/1 ÷ 1/1

<2%$+

=2754$(*+

>"()7$+

Photo courtesy of Peter Lee, FSU

Photo courtesy of J. Minervini MIT

Φ = 0.81 mm

Φ ∼ 50-100 µm

Φ ∼ few µm

Multi-filamentary superconducting wires


!"#$%&'()"&*(+,-.%$/,0/1%2()/34,

,

! ,15.(,/6&6,7829$(1/,09"8*:7829$(12%;3<,

! /6&6,7829$(1/,-.15.(,2,=",921%.><,

! ,1-./1$),7829$(1,/1%"&1"%$6,

The ITER strands


Fabrication of s.c. wires


NbTi

Fabrication of s.c. wires


Nb3Sn

!"#$%&'()"*+)

,-)."/&0&1)!"#$%#&'()"*

+%,(-'#."/*.(*0123(*

3,4"#)%(-,)5%#/*234456)

It requires a heat treament at 650 °C to form the s.c. phase.

Once formed, it is a brittle material!

Practical Materials


Practical Materials


NbTi

Nb3Sn

Bpeak determines the choice of the s.c. material to be used

ITER TF

ITER PF

Superconducting cables


How should a certain number of s.c. wires be assembled into a

cabled structure, that constitutes the conductor, by which fusion

coils are wound?

Cryogenics of superconducting magnets


Typically: T ~ 4.5K, P ~ 6 ÷10 bar

Steel

jacket

Empty fraction for forced He

circulation

1000 ÷ 5000

filaments

strand

Pressure relief channel

Strand bundle

qS−He = S ⋅hHe−S (TS −THe )

Convective heat transfer:

BUT: Low JENG

-  Effective cooling

-  Mechanically strong

-  Flexible layout

-  Effective electrical

insulation

Hoenig; Montgomery; Iwasa (1975):

high cooling efficiency of single phase (supercritical) He in turbulent flow

and in direct contact with a large wetted surface

Outline




conductors (CICCs)



!  DEMO



Manufacturing of ITER Conductors


Manufacturing of ITER TF Coils: winding


Conductor is wound in Double-Pancakes

Manufacturing of ITER TF Coils: heat treatment


Large furnaces required for curing Nb3Sn for about 3

weeks, in temperature steps up to 650 °C +/- 5 °C

Manufacturing of ITER TF Coils: radial plates


ITER TF Radial Plate

Radial plate flatness: 1 mm

Manufacturing of ITER TF Coils: closure welding


2 Nd–YAG laser welding 3 robots; 1.5 km weld per pancake

Manufacturing of ITER TF Coils: final insulation


ITER Coils: a technological challenge … won!


Outline




conductors (CICCs)



!  DEMO



From ITER to (EU)-DEMO


DEMO CAD model (30 April 2014)

www.euro-fusion.org/eurofusion/roadmap/ DEMOnstrate: 500 MW

electric power

TF magnet ITER July 2013 April 2015 July 2018

Major radius (m) 6.2 9 9.072 9.073

Toroidal field on axis (T) 5.2 6.8 5.7 5.3

Plasma current (MA) 9.1 14 19.6 17.9

Number of TF coils 18 16 18 16

TF magnet ITOT (MA) 164 305.8 257.1 238.7

Stored energy/TF coil (GJ) 2.28 9.07 7.54 10.04


30

2014

2014 2018

16 TF coils

ITFcoil=14.9 MA

2015 2015 18 TF coils

ITFcoil=14.3 MA

Bpeak=12.2 T

2013 16 TF coils

ITFcoil=19.1 MA

Bpeak=13.3 T

!  since 2011 R&D activities in EU on the magnet system


TF magnet ITER July 2013 April 2015 July 2018

Major radius (m) 6.2 9 9.072 9.073

Toroidal field on axis (T) 5.2 6.8 5.7 5.3

Plasma current (MA) 9.1 14 19.6 17.9

Number of TF coils 18 16 18 16

TF magnet ITOT (MA) 164 305.8 257.1 238.7

Stored energy/TF coil (GJ) 2.28 9.07 7.54 10.04

CICC Operating Current 68 kA 70 – 95 kA

CICC Peak Magn. Field 12 T 11.5 – 13.5 T

E.m. Load (kN/m) 816 800 – 1200

∆Tmargin 0.7 K 1.5 – 2 K



From ITER to DEMO,

CICC design is required to be:

-  With optimized performance;

-  With stable behavior;

-  Cost-effective!

!  Quite some effort is being spent

on technology development

"  since 2011 R&D activities in EU on the magnet system


32

• Degradation

with e-m cycles

• ∆Tmarg > 1.3K

@ 68kA,12T

•  Total strain

εeff=-0.70%

ITER TF

!  Higher Ic

!  Less superconducting strands

• ∆Tmarg >2K @ 82kA,13T

• εeff ∈[-0.55, -0.35]%

DEMO TF conductors


Outline




conductors (CICCs)



!  DEMO



The Italian DTT project


(Sept. 2018) www.euro-fusion.org/eurofusion/roadmap/

DTT: Divertor Tokamak Test Facility project


1.  Divertor concept; 2.  Magnetic configurations

3.  Liquid metal plasma facing components

a 500 M€ Italian project, being built

in Frascati

General objective: create a

research infrastructure addressed

to the solution of the power

exhaust issues in view of DEMO.

Test Divertor alternative

solutions & improve

experimental knowledge in the

PEX scientific area

D-shaped superconducting tokamaks and DTT

6.2m

EAST (A=3.5 T, Ip=1 MA)

KSTAR (A=3.5 T, Ip=2 MA) 1.8m

1.1m

SST-1 (A=5.5 T, Ip=0.22 MA)

JT-60SA(B=2.25 T, Ip=5.5 MA)

3.1m

1.7m

ITER (B=5.3 T, Ip=15 MA)

10.0 kA - 5.1 T NbTi

14.3 kA - 5.8 T NbTi

35.2 kA - 7.2 T

Nb3Sn

68.0 kA - 12 T Nb3Sn

25.7 kA - 5.7 T NbTi

∼ 45 kA –!11.7 T Nb3Sn

2.1m DTT

6 PF coils

Identical in pairs to guarantee full

top/down symmetry

NbTi (PF2 to PF5) CICC: 28.6 KA – 5.4 T

Nb3Sn (PF1 & PF6) CICC: 28.3 KA – 9.1 T

DTT Magnet System

37


18 TF coils:

Nb3Sn CICC: 44.8 KA – 11.9 T

providing 6.0 T over plasma

major radius (2.14 m)

6 CS modules

Nb3Sn CICC: 29 KA – 13.4 T

providing 16.2 Weber magnetic flux for

plasma initiation at breakdown

PF1

PF2

PF3

PF4

PF5 PF6

CSU3

CSU2

CSU1

CSL3

CSL2

CSL1

CRYOSTAT

VACUUM VESSEL

GRAVITY SUPPORT

SC FEEDERS

IV COILS

DTT parameters and exploitation plan


•  !"#$%&'"()"%*+,,-./0%!"!#$%$%

1.2+*34-%5&-,"35.%67/8%"%)"#$%

8798%β%:"/%/5&%+&%/5%!&#'$%

•  ;-"3.9%(<(/-)%&,54727.9%(!#

$)#=/",9-/>%=?@AB@%!C%DEF;G%

BAH%!C%1EF;G%IAJK%!C%LLM1>$%

R (m) / a(m) 2.14/0.65 SN – 2.14/0.65

DN

A 3.3 SN/DN

Vol (m3) ≈28

Ip (MA) 5.5

BT (T) 6 @ R0

Neutron production rate, Sn (n/s) 1.2-1.5 10^17 DD + 1%

DT

Maximum dwell time for high performance

3600

Nominal repetition time after disruption (s)

3600

Number of shots per day 5-10

Days of operation per year 100

Years of operation 25

Outline




conductors (CICCs)



!  DEMO



HTS in fusion COILS

40

Why use HTS in fusion?


ENABLING or EXTENDING technology?

Substitute or Integrate LTS?

What’s the real goal?

Why HTS in fusion? (wrt LTS)

41

1.  Possibility to work at higher temperature (10 ÷ 20 K , or beyond).

2.  Possibility to work at high Iop/Ic, with large temperature margins.

3.   Possibility to access higher fields (> 15 T ?)

! TF: highly beneficial for plasma confinement and fusion power

Stability quantified by the parameter:

β =Gas!Pr essure

Magnetic!Field !Pr essure!

=p

B22µ

0

=n!kBT

B22µ

0

Fusion Power: PFUS

∝β 2 ⋅B4www.tokamakenergy.co.uk

psfc.mit.edu/sparc



42






β =Gas!Pr essure


=p

B22µ

0

=n!kBT

B22µ

0

Fusion Power: PFUS

∝β 2 ⋅B4

“the possibility to access B > 20 T would be a

“game changer” for fusion development”

Affordable, Reliable, Compact

(D. Whyte; J Fus Energy 2016)

www.tokamakenergy.co.uk

psfc.mit.edu/sparc



43






β =Gas!Pr essure


=p

B22µ

0

=n!kBT

B22µ

0

Fusion Power: PFUS

∝β 2 ⋅B4

“the possibility to access B > 20 T would be a

“game changer” for fusion development”

Affordable, Reliable, Compact

(D. Whyte; J Fus Energy 2016)

www.tokamakenergy.co.uk

psfc.mit.edu/sparc



44





! CS coils:

Higher Magnetic Field

Smaller size (e.g. smaller outer

radius at same magnetic flux!)

reduced size and cost

of the whole tokamak

Wesche_IEEE TAS 16



An emerging paradigm ….

45

NOT necessarily:

- public-funded

- tens of billion-size

- large projects

- requiring large organizations ITER site; gen. 2015



46

NOT necessarily:

- public-funded

- tens of billion-size

- large projects

- requiring large organizations

But also:

- private-funded

- hundreds million-size

- compact-machines

- managed by medium-size

start-ups

MIT/CFS ARC Reactor studies



Tokamak Energy (UK) Commonwealth Fusion Systems (USA)

Tri-Alpha Energy (USA) General Fusion (Canada)



Tokamak Energy (UK) Commonwealth Fusion Systems (USA)

Tri-Alpha Energy (USA) General Fusion (Canada)

HTS S.c. Tokamaks

Spherical T.

? LTS ? technology

Reversed Field Pinch configuration

NON s.c. fusion

HTS in DTT?

49 L. Muzzi - Seminario Roma TRE - 5 Dic. 2019 49

Design of a HTS insert

(presently outside of the DTT scope!)

Pre-compression structures

Coil centering systems

HTS in DTT?

50 L. Muzzi - Seminario Roma TRE - 5 Dic. 2019 50

Design of a HTS insert (presently outside of the DTT scope!)

Pre-compression structures

Coil centering systems

Technical merits:

-  Higher Magnetic Flux (16.2 Volt sec !

17.3 Volt sec)

-  Higher Magnetic Field (13 T ! 16.5 T)

-  Fundamental technology

demonstration toward high-magnetic

field fusion.

HTS in DTT?

L. Muzzi - Seminario Roma TRE - 5 Dic. 2019 51

Exte rna l h igh -s t reng th Aluminum-alloy jacket.

6 slots for twisted stack of

s.c. tapes

30 – 35 kA current in 18 T –

20 T field (to be tested)

ENEA-TRATOS Aluminum slotted Core HTS CICC

Summary and conclusions


-  Superconducting coils are (luckily) an “enabling technology” for

nuclear fusion!

-  Present (or near future) reactors are based on LTS CICCs: fairly

mature technology, even though still with optimization margins.

-  There is some potentiality for HTS in this field, to integrate (not

substitute!) LTS technology or to open new routes.

-  A new paradigm is emerging: from large, public-funded projects, to

smaller scale, high-field, compact tokamaks, managed by medium-

size start-ups. For many of these, the development of HTS

technologies represent the breakthrough.

Luigi Muzzi

[email protected]


Documents

Conductors and Coils for nuclear fusion magnets: from state-of … · 2019-12-17 · L. Muzzi - Seminario Roma TRE - 5 Dic. 2019 17 NbTi Nb 3Sn B peak determines the choice of the