Progress Towards Compact Fusion Reactor · Fusion Reactor does not need to be big! A recent study*...

Progress Towards Compact Fusion ReactorMikhail Gryaznevich

27 May 2015

8th IAEA Technical Meeting on Steady State Operation of Magnetic Fusion Devices, 26‐29 May 2015, Nara, Japan

• Who we are and what are our goals• New high field ST, under construction• Steady-state operation issues on the

path to compact Fusion Reactor and the role of ST40

• Technology development at TE Ltd

• First plasma in ST40 is expected in early 2016

Who are we?

• Tokamak Energy Ltd is a private company funded by private investors, it is based at Culham, UK

• Two small tokamaks are operating at Tokamak Energy Ltd Milton Park site: Cu ST25 and full‐HTS ST, third tokamak under construction, 3T ST40

Who are we?

What is our goal? TE Ltd Path to Fusion

• Our goal is demonstration of the possibility of electricity production in a fusion reactor based on high field ST

Conventional path

TE path

TE Path to Fusion, why high field ST?

Fusion powerFusion power VolumeVolumeEfficiencyEfficiency Field strengthField strength

• Engineering of high field in ST is a real challenge! Physics is good.

• Increase in beta allows significant reduction in plasma volume

TE Path to Fusion, why high field ST?

Apparently the high beta potential of the ST is so great that the physics of this device will not determine its size.

Ron Stambaugh, “THE SPHERICAL TOKAMAK PATH TO FUSION POWER”, FUSION TECHNOLOGY VOL. 33 JAN. 1998

Fusion Reactor does not need to be big!A recent study* has shown that Qfus depends mainly on Pfus and the H factor, and only weakly on device size:

At high field, high H factor is expected in STs opening the possibility of low power, relatively small, high Qfus devices.

*A E Costley, J Hugill and P Buxton, Nucl. Fusion 55 (2015) 033001

ST40 – a new high field ST, under construction

• First plasma expected in early 2016

ST40 project

• To be constructed in 2015 at Milton Park TE Ltd site

Main features:• High toroidal field up to 3T, water or LN2 cooled copper magnet

• Plasma major radius 0.4 ‐0.6m, R/a = 1.6‐1.8

• Moderate elongation (~2.5) and triangularity(~0.3), DND

• Plasma current up to 2MA, merging/compression plasma formation

• NBI and EBW/ECRH heating• Possibility of DT ops

Why we are building ST40?

• High field in ST is a real challenge! That is why we need to build a new high field ST.

Water cooled magnets

First plasmasIp=0.5MA Bt=1T

Power Supply is based on commercial MAXWELL supercapacitormodules and capacitor banks

Next phase: LN2 COOLED TF magnetHigh performance

Ip=2MA Bt=3T

ST40 is a short pulse device

• ST40 is a step to steady‐state high field STs

Work packages

• Coils and coil support structure

• Critical path ‐ current focus

• Inner Vacuum Chamber• Near critical path ‐

current focus

• Control system

• Vertical stability system

• Diagnostics

• Power supply

• Safety system

3D model of central column wedges complete and requests for quotation issued

Parts for current return limb joint test rig ordered

TF coil shape, position and restraints fixed

PF coil positions specified pending mechanical review

Geometric constraints on angled NBI ports identified. Optimum within these being identified

Passive stability plate geometry determined pending protection review

ST40 construction, present status

• TE Ltd has an experienced management team

Objectives of ST40, and Physics Programme issues

• We are not setting ambitious targets for plasma parameters at this stage (high temperatures, betas, densities etc.), just demonstrating that the tokamak is operating.

ST40 Objectives for Day‐1:• Construction and commissioning device for agreed specification

‐ all systems must be built and commissioned, or designed. • Physics objectives include:

‐ demonstration of first plasma‐ achievement of target operation parameters: 0.5MA, 1T in 50‐120ms pulses‐ demonstration of Day‐1 diagnostics and data&controlsystems


• These objectives should be achieved without major modifications

Main objectives of ST40 after Day‐1 objectives are met:• Demonstration of advantages of high field in ST

‐ achievement of target operation parameters‐ demonstration of high quality plasmas: high temperatures & density, low Zeff‐ demonstration of improvement of performance with field, leading to high confinement, high bootstrap, high beta etc.


• These objectives require major upgrades of power supplies and heating systems

• Operations in DT will require moving to another site

Next stage ‐ utilisation of advantages of high field in ST:‐ demonstration of feasibility of CFNS (not in steady state)‐ demonstration of possibility of Qfus~ 1 in a driven regime (i.e. 1‐2MW of Fusion power using 1‐2MW of NBI or other auxiliary heating, EBW/ECRH considered)‐ both would benefit from DT ops, with DD equivalent demo first

Neutron production and demonstration of Qfus~1

• High performance (temperatures in 5‐10keV range) can be expected even with ohmic heating and even with cautious assumptions for energy confinement time

• ST40 can be a prototype of the most efficient and intensive Neutron Source, producing up to 1018n/s in DT

• If confinement improves as predicted by theory and by MAST/NSTX experiments, then achievement of Qfus~ 1 in ST40 (in a driven DT regime) is not impossible

• These objectives require major upgrades of power supplies and heating systems

• Operations in DT will require moving to another site

Summary of ST40 objectives

• Main engineering objective:‐ to demonstrate feasibility of high‐field ST

• Main physics objective:‐ to demonstrate advantages of high field in ST, aiming at 10 keV‐range plasma temperatures

• Main objectives of possible upgrades:‐ demonstration of efficient production of neutrons ‐ demonstration of Qfus~1 with upgraded NBI system in DD equivalent and with DT subject to site availability and tritium licence.

• ST40 will be the first high field Spherical Tokamak• Necessary step on the way to steady‐state HTS ST

Plasma formation and current ramp

• m/c has been proven on START, MAST and other STs

• Coils for merging/compression will be installed inside the vacuum vessel.

• M/c is expected to provide up to 2 MA plasma current• Small central solenoid (~0.2Vs) will be used at flat‐top

What plasma parameters can be achieved

• Heating method strongly affects performance.

ASTRA+NUBEAM simulations: R=0.4m, A=1.8, k=2.5, δ =0.4, Ipl= 2MA, Bo=3T, <ne>=1·1020m-3, Zeff=2, Zimp=6, OH or 1MW additional heating. D-D ops.• For additional heating different

points assume three cases:- all power going into ions- all power going into electrons- 50% to ions, 50% to electrons

• High temperatures can be expected even in OH ops and even at H98=1, to reach high performance at 3T ~0.5s will be needed.

• The required additional heating power may be higher, depending on losses and deposition efficiency.

More ASTRA simulations for ST40

• Details in following back‐up slides

• High neoclassical resistivity in STs, so good ohmic heating, MW level

• Uloop ~0.5‐0.1 V, ~1/H, so good use of small solenoid at flat‐top

• Confinement time in sec range can be expected in OH regime

• Auxiliary heating ~1MW weakens τE dependence on H98• Ion heating is more efficient (choice of heating method)• Auxiliary heating at low H98 is not efficient

What plasma parameters can be achieved

• ST40 will be the first Spherical Tokamak to test these predictions at high TF required in ST reactor

R=0.4m, A=1.8 k=2.5, δ =0.4, Ipl= 2MA, Bo=3T, <ne>=1·1020m-3, Zeff=2, Zimp=6 OH or 1MW additional heating. D-D ops.

• Ion thermal transport in STs has been shown to be close to neoclassical

• MAST/NSTX confinement scaling predicts very high H‐factors at high TF and electron thermal transport close to neoclassical

ASTRA modelling and experimental results from MAST, NSTX and Globus‐M suggest at least linear increase in confinement with TF


• OH is very efficient in spherical tokamaks

OH regime. Bt =3T, Ipl=2MA, ne=1020m‐3, Zeff=2, H98=0.25 ‐ 2

0 1 2 3 4 5 6 70

1

2

3

4

5

6

H98=2

H98=1

H98=0.5

P OH, M

W

Te,keV

~T-1.35e

H98=0.25

• High neoclassical resistivity in STs

0 1 2 3 40

1

2

3

H98

~1/HU, V


• Small solenoid will be useful at flat‐top

0 1 2 3 40,00,20,40,60,81,01,21,4

H98

~H2.1 , s

• Confinement time in sec range can be expected in OH regime• Will pulse duration be enough to demonstrate high confinement?

0 1 2 3 40,00

0,05

0,10

0,15

H98

~H3/4

U*

E, Vs

OH regime. Bt =3T, Ipl=2MA, ne=1020m‐3, Zeff=2, H98=0.25 ‐ 2


• Choice of auxiliary heating

• Auxiliary heating 1MW weakens τE dependence on H98• Ion heating is more efficient• Heating at low H98 is not efficient

Additional heating regimesBt =3T, Ipl=2MA, ne=1020m‐3, H98=2, 1MW to ions or electrons

1 2 3 4 5 60,0

0,5

1,0

1,5

2,0

~H2

E, s

H98

~H1.2

Additional P=1MW

OH

1 2 3 4 5 6

5

10

15

20

~H0.5Te, k

eVH98

~H0.9

1MW to electrons

OH

1 2 3 4 5 6

5

10

15

20

~H0.5

Ti, k

eV

H98

~H0.8

1MW to ions

OH

Other design issues:- vertical stability- divertor

• Vertical stability, divertor load are crucial issues

• High elongation and betas are required for s/s

• Non‐solenoid plasma formation, current ramp and sustainment, divertor and wall loads

• High field ST allows flexibility in the choice of current drive methods

Towards steady-state operations

• ST approach can provide very high bootstrap current• High bootstrap requires high elongation (STs have high natural elongation) and beta (STs achieved record betas)

• As elongation will be much above the natural one, feedback system and passive stabilisation will be required

Vertical stability and passive plates

• Growth rates < 100 s‐1 are achievable, KINX, S Medvedev

• Elongations in ST40 will be well above “natural”, so both passive and active stabilisation is needed

• Growth rate is very sensitive to position of passive plates (KINX simulations, 2MA, 1MA and 0.5MA cases are shown)

• Internal coils for active feedback will be used

Vertical stability and passive plates

= 41 s‐1

Cu 2cm 2cm thick steel wall

= 58 s‐1unstable

• The upper divertor is not very efficient: = 41 s‐1 ‐‐> 51 s‐1

• But the conductors close to the X‐point are essential to hold the plasma tips: = 51 s‐1 ‐‐> 81 s‐1

Cu 2cm + vessel No upper plate

= 51 s‐1

• Growth rates < 100 s‐1 are achievable

Crucial importance of both inner and outer passive plates

• Only inner or outer passive plates are not enough• Passive plates can be LN2 cooled, so even lower growth rates

• Growth rates < 50 s‐1 are achievable with inner+outer passive plates

Divertor design

• Divertor does not require active water cooling and divertor loads are moderate

• Each divertor is divided in 32 components made of a CFC elongated tile for the PFC and a 316L stainless steel support bolted to the vacuum chamber.

Divertor design


Target power load estimated with Eich formula:- According to NSTX scaling: λq ~ 1 cm in ST-40 with Ip = 1MA- According to JET & ASDEX scaling: S ~ λq/4 (worst case)- Assumption: PIT = 0.1Pheat & POT = 0.9Pheat

• Peak heat flux density q┴IT = 1 MW.m-2 & q┴OT = 5 MW.m-2

• Angle of incidence on target ~ 3o

Divertor design


• Temperature distribution estimated with following assumptions: - semi-infinite model with Tsurf = 2∆tq┴ / (ρCpKπ)0.5

- conservation of heat in 1D- temperature distribution in depth ~ exp(-x2)

• According to this conservative model, after 2 s discharge:- CFC surface temperature < 1000 C- Interface CFC / 316L SS temperature < 100 C

T, 0C

Optimisation of Neutral Beam Current Drive and Heating:

- what shall we optimize?

• Importance of full‐orbit simulations

Heating optimisation in ST

• High density in ST40 helps to reduce shinethrough

• Optimisation of heating aims to deposit beam where required: on-axis, off-axis etc., and to reduce fast ion losses.

• Shine through can be avoided by consistent choice of beam energy and deposition chord length with the target density

• To reduce fast ion losses, beams should not be deposited too close to the plasma edge. To minimise losses, the most of deposition should within the limiting orbit

Variation of possible injection chords

Optimisation of CD in “driven” CFNS regime

• CD optimisation strongly depends on plasma current

Ipl =1.5MA

1.0 1.5 2.0 2.5 3.0

0

10

20

30

40

50

Opt

imal

ver

tical

ang

le

Plasma current, MA

• With currents of 1-3MA and average density 2x1020m-3 beam losses on shine through, on orbits and on charge exchange are small

• Increase in the NBCD with the beam energy is significant, but shine through also increases at lower densities (more details later)

• There is an optimum beam injection vertical angle depending on Ipl when hot particle trajectories are located predominantly along the field lines.

PNBI~5MW, R=0.5m, A=1.6, k=2.75, Bt=1.5T, DT, ASTRA+NUBEAM

Optimisation of CD in ST40

• Conclusion: NBI is not efficient for direct CD in ST40

• Strong dependence on vertical injection angle, and for toroidal angle• Lower beam energy results in lower efficiency• Gain from optimisation is very significant, >100%, but poor CD

efficiency even for best optimisation• For burning ST, optimisation for torque is more efficient, and easier,

than optimisation for direct CD

R/a=0.4/0.22m, k=2.5, Bt=3T, Ipl=2MA, PNBI=1MW, Eb=40kV, D‐D

Optimisation of CD in ST reactor

• It is easier to rotate ST plasma compared with a conventional tokamak and momentum confinement may be up to 5 times higher than the energy confinement (NSTX)

• Optimisation of torque may be more efficient to increase CD in high field ST reactor:

- in STs, confinement improves with plasma rotation, MAST results (G Counsell, 20th FEC):

• Rotation is provided by NBI torque, PNBI ~ 2MW, ~ 50kV, DD.

Optimisation of CD in ST40

• Beams with different energy and launch are needed• Torque is maximised using low energy beams (ASCOT)

• Optimisation for torque results in different optimal launch geometry for ST40 (PNBI=1MW, Eb=40kV, DD, may differ for D‐T and T‐D):

‐much smaller poloidal angle‐ less dependent on toroidal angle

Optimisation of CD in ST40, ASCOT

• Simulations with ASCOT confirm NUBEAM results

• Two possible positions for NBI port.

• Many possible injection angles.

• Shinethrough issues

Upper port launchMid‐plane port launch


• Beams with different energy and launch are needed

Mid‐plane

port lau

nch

Upp

er port lau

nchOptimum angle

for torque

Optimum angle for current drive

Significant shine through

1MW, 70keV


• Lower energy NBI and midplane launch for torque?

1MW, 45keV

1M, 120keV

More torquefrom lower energy

More currentfrom higher energy


• Taking NBI efficiency into account reverses results!

2MW, 45keV

5MW, 120keV

More torque from higher energy

More current from higher energy

Taking into account for relation between NBI energy and NBI power

-particles containment & wall load in ST Pilot plant

• Full‐orbit simulations show possibility of reducing Ipin compact ST reactor

• Monte‐Carlo Beam Deposition Code NFREYA and Fast Ion Fokker‐Planck Code FIFPC, comparison with guiding centre model

Tokamak Energy Ltd small tokamaks: - mock-ups and prototypes towards steady-state operations

• Two small tokamaks operational, research on EBW CD and High Temperature Superconductors

Tokamak Energy Tokamak Engineering Centre

• Two small tokamaks are operating at Milton Park site: Cu ST25 and full‐HTS ST

Small tokamaks

• Two small tokamaks are operating at Milton Park site: Cu ST25 and full‐HTS ST

Main parameters:‐ Present circular section vessel:R, a = 25 / 12.5cm‐ Aspect ratio = 2‐ Toroidal field 0.1 – 0.3T‐ ST25 Pulse length (RF plasmas): ~30s‐ RF: 2.45GHz, 3kW CW, Homer autotuner‐ Solenoid: 4‐layer, 348 turns, 100mVs for 3kA to ‐3kA swing, Ip ~10kA‐ Power supplies: capacitor banksSlow: Maxwell 125V, 63F modulesFast: Maxwell 11kV 0.5mF

‐ HTS‐ST pulse length (RF plasma): 24h‐ RF: 13 MHz < 1kW

ST25

HTS‐ST

Tokamaks with HTS Magnets

• First use of HTS in tokamak magnets

Spot the difference: GOLEM August 2011 left, ST25 October 2013 right

• First tests of HTS coils on ST25 during 2.5 sec plasma discharge. 4 October 2013.

• Coil temperature 8‐20K. Current in tape 200‐400A.

Joined efforts: Technical University of Prague, IPP Prague, Tokamak Solutions UK, Oxford Instruments, Forma Machinery LV

Continuous discharge in fully-HTS tokamak

• First 24h pulse completed on October 30 2014

First plasma in full-HTS tokamak, Oxford Instruments and Tokamak Energy Ltd, July 2014, 100 sec pulses

RF 24h discharges on HTS-ST25

• World record plasma durations in a tokamak

13 MHz RF antenna assembly

Plasma in HTS‐ST• Operations in He

and H• Next stage: high

plasma current operations

EBW experiments on ST25, simulations

• ST25 is the only tokamak with HFS and fully adjustable LFS EBW launch

• LFS and HFS launch is used. Simulations for f=2.45GHz, Bt(0)=0.09T, ne(0)=1.5x1017m-3, Te(0)=0.1keV, Zeff=1

LFS launch

HFS launch

• HFS launch is up to 6 times more efficient than LFS launch, expected ~1.8 kA for 2.0 kW

EBW experiments on ST25, results

• Convincing demonstration of EBW current drive

• Plasma current changes direction depending on the angle of antenna (LFS launch)

• HFS launch: green – no BV, red –positive BV, blue – negative BV

• Longest pulses up to 28 seconds

CONCLUSIONS

• ST40 will be the first high field Spherical Tokamak

• Demonstration of burning plasma is the current challenge for Fusion and can be achieved in a compact high-field ST

• The ST path to commercial application of Fusion can start from compact ST with R as low as 0.4 m with support of world-wide Fusion research

• Innovations can make Fusion sooner and cheaper

• Non-inductive current drive and start-up are crucial for steady-state operations and can be demonstrated in a compact high-field ST40

Progress Towards Compact Fusion Reactor · Fusion Reactor does not need to be big! A recent study*...

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