Studies of Next-Step Spherical
Tokamaks Using High-Temperature
Superconductors
Jonathan Menard (PPPL)
22nd Topical Meeting on the Technology of Fusion Energy (TOFE)
Philadelphia, PA
August 22-25, 2016

Multi-institutional effort exploring
low aspect ratio tokamak concepts

Paper summarizing 5 year study of Cu and HTS STs recently published in Nuclear Fusion

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Possible missions for next-steps
1. Integrate high-performance, steady-state, exhaust Divertor test-tokamak - DTT

2. Fusion-relevant neutron wall loading Γn ~ 1-2MW/m2, fluence: ≥ 6MW-yr/m2

3. Tritium self-sufficiency
 Tritium breeding ratio TBR ≥ 1

4. Electrical self-sufficiency
 Qeng = Pelectric / Pconsumed ~ 1

5. Large net electricity generation

This talk will
discuss PPPLled studies of
how low-A“spherical”
tokamaks
could fulfill
these missions

 Qeng >> 1, Pelectric = 0.5-1 GWe
3What is optimal A for HTS FNSF / Pilot Plant?
• Pfus / V ~ ε(βN κ BT )4 at fixed bootstrap fraction• βN and κ increase at lower aspect ratio• But BT decreases at lower A – depends strongly on:– Thickness of inboard shielding and breeding blanket– HTS allowable field and current density

4Approach:• Fix plasma major radius and heating power (50MW)– R0 = 3m – smallest size for Qeng > 1 and high fluence• Apply magnet & plasma constraints (see backup)– HTS strain: 0.3%, βN: n=1 no-wall, κ: 0.95×limit, fGW = 0.8• Vary aspect ratio from A = 1.6 to 4
• Vary HTS current density, peak field– Also scan inboard shielding thickness• Compute QDT, Qeng, and required H98 (unconstrained)
5HTS cables using REBCO tapes achieving
high winding pack current density at high BT
10 mm

Conductor on Round
Core Cables (CORC)
JWP ~ 70MA/m2 19T

10 mm

Base cable: 50 tapes YBCO Tapes with 38 µm substrate(Van Der Laan, HTS4Fusion, 2015)

Higher Jcable HTS
cable concepts
under development:

Base Conductor
He Gas Cooled
8kA,
JWP ~ 160MA/m2

7 kA CORC (4.2K, 19 T) cable
7 mm

Copper
7 mm

SS conductor
jacket for strength6

High TF winding-pack current density
required to access highest BT at lower AJWP

[MA/m2]

19T: Present
CORC HTS limit

Max BT at TF coil [T]20181614

12T: ITER-like
TF coil limit(Nb3Sn, 11.8T)• Coil structure sized to maintain ≤0.3% strain on winding pack for
all cases shown here
• Effective inboard WC neutron
shield thickness = 60cm

12108
1.5
987
65
43
2
10
1.5

2.5
3.0
3.5
BT2.0
at geometriccenter
[T]
A

2.0

2.5

A 3.0

3.5

4.0

4.0

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High current density HTS cable motivatesconsideration of low-A tokamak pilot plants J• ITER-like TF constraints:– JWP=20MA/m2, Bmax ≤ 12T– Pfusion ≤ 130MW– Pnet < -90MW• JWP~ 30MA/m2, Bmax ≤ 19T– Pfusion ~ 400MW– Small Pnet at A=2.2-3.5• JWP ≥ 70MA/m2,Bmax ≤ 19T– Pfusion ~500-600MW– Pnet = 80-100MW at A=1.9-2.3Pfusion [MW]800

WP[MA/m2]

600(12T)

400
2000
1.5
150

2.0

2.5
P

2.0

2.5

3.0

3.5

4.0

A 3.0

3.5

4.0

net [MWe]A

100500-50-100-150

A ~ 2 attractive at high JWP

1.58A ≥ 2 pilot plant scenarios have elevatedH > 1, IP = 6-12MA, fBS ~ 80% (not shown)
JWP[MA/m2]

Effective inboard WC n-shield thickness = 60cm

9A ≤ 2 maximizes TF magnet utilization
Eff. shieldthickness:
0.3m
0.4m
0.5m
0.6m
0.7m

JWP = 70MA/m2

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Need inboard breeding for TBR > 1 at A=2TBR
Required
TBR = 1.04
(4% margin)

Outboard
TBR ~ 0.92
IB blanket thickness [cm]

10cm IB blanket sufficient for TBR > 1.04
11Model blanket and shield thickness vs. A

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A ≥ 3 maximizes blanket utilization
Eff. shieldthickness:
0.3m
0.4m
0.5m
0.6m
0.7m

JWP = 70MA/m2

13A=2, R0 = 3m HTS-TF FNSF / Pilot Plant
BT = 4T, IP = 12.5MA
κ = 2.5, δ = 0.55
βN = 4.2, βT = 9%
H98 = 1.8, HPetty-08 = 1.3
fgw = 0.80, fBS = 0.76
Startup IP (OH) ~ 2MA
JWP = 70MA/m2
BT-max = 17.5T
No joints in TF
Vertical maintenancePfusion = 520 MW
PNBI = 50 MW, ENBI = 0.5MeV
QDT = 10.4
Qeng = 1.35
Pnet = 73 MW

Cryostat volume ~ 1/3 of ITER〈Wn〉 = 1.3 MW/m2
Peak n-flux = 2.4 MW/m2
Peak n-fluence = 7 MWy/m2
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Inboard and outboard blanket
vertical maintenanceOutboard blanketremoved

Inboard blanket removed once outboard
blanket sectors removed – depending on
the toroidal extent of the inboard blanket

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Long-leg divertor

qdiv = 9MW/m2, Rstrike = 4.2m

Liquid metal (short-leg)
qdiv = 21MW/m2, Rstrike = 2.5m

Detachment can further reduce qdiv ~ 2-3x
Long-leg / Super-X
Geometry also compatible
with “vapor box” concepts(Ono / Goldston)

Shorterleg LMdivertor

16Exploring liquid metal divertor design
similar to flowing water curtain systems

LM injector system canbe assembled in asingle or double unit

LM containment structure
Shield block
Ferritic steel backing plate
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HTS ST-FNSF design with Li flow on
divertor and inboard surfaces
Double null liquid
metal divertor system
Li flows from upper
divertor down the inboard
wall, exiting just after thelower inboard divertor.
Separate Li cooling oflower divertor

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Benefits of shorter-leg LM high-heat-flux divertor:• Significantly reduce outboard PF coil current– Reduced PF size, force, structure• Eliminate separate upper cryo-stat (for PF5U)

45326178• Li wall pumping could help increase H19

Summary• Developed new self-consistent configurations forlow-A FNSFs / Pilot Plants– Long-leg and/or LM divertor, T self-sufficient, only exvessel TF and PF coils, vertical maintenance• Compact Pilot Plants achievable by combining
improved stability of low-A + advanced magnets– Optimal A will be informed by results from NSTX-U andMAST-U and REBCO TF magnet development• Liquid metal divertors for high heat flux could
simplify cryostat, reduce coil currents/forces– Higher confinement from liquid Li also beneficial20