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Compact Fusion Energy based on the Spherical Tokamak
A. Sykes1, A. E. Costley1, C. Windsor1, O. Asunta1, G. Brittles1, P. Buxton1, V. Chuyanov1,
J.W. Connor1, M.P. Gryaznevich1, B. Huang1, J. Hugill1, A. Kukushkin1, D. Kingham1,
J.G. Morgan2, P. Noonan1, J.S.H. Ross1, V. Shevchenko1, R. Slade1, G. Smith1,3
1Tokamak Energy Ltd, Culham Science Centre, Abingdon, OX14 3DB UK

2

Culham Electromagnetics Ltd, Culham Science Centre, Abingdon, OX14 3DB

3

Department of Materials, University of Oxford 16 Parks Road, Oxford OX1 3PH UK

Email contact of main author: alan.sykes@tokamakenergy.co.uk
Abstract. Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor
magnets as a possible route to fusion power using relatively small devices. We present an overview of thedevelopment programme including details of the enabling technologies, the key modelling methods and results,
and the remaining challenges on the path to compact fusion.

1. Introduction
Since the mid 1980s the spherical tokamak (ST) has been recognized as an important device
for fusion research [1,2,3]. Such devices demonstrate all the main features of high aspect ratio
tokamaks but are relatively small and inexpensive to construct. Moreover, research has
shown that they have beneficial properties such as operation at high beta, lower H mode
power threshold, reduced tendency to disrupt, and possibly higher confinement. Earlyattempts to design reactors based on STs, however, did not produce convincing designs, anduntil recently STs have been mainly seen as useful research devices and possibly as neutron
sources for component testing. However, recent advances in both tokamak physics andsuperconductor technology have changed the situation, and relatively small STs operating athigh fusion gain may now be possible. The key physics step is the realization that the power
needed for high fusion gain may be considerably less than previous estimates, while the key
technological step is the advent of ReBCO high temperature superconductors (HTS). In
addition to operating at relatively high temperatures, HTS can also produce and withstandrelatively high magnetic fields: both of these properties are beneficial in the design ofmagnets for fusion devices especially for STs where space is limited in the central column.
Utilising these developments, Tokamak Energy (TE) is developing an alternative route tofusion power based on spherical tokamaks constructed using HTS magnets, and modellingand concept work is underway to determine the optimum power and size of an ST/HTS fusion
module. This work identifies key aspects in the physics and technology that significantly
affect the size, power and feasibility of such a module. In parallel, experimental work isunderway addressing these aspects, including the construction and operation of a series ofSTs. In this paper, we present recent new results and the status of the development
programme, and we outline the intended next steps.
The paper is divided into five main sections. In section 2 we summarise briefly our earlier
modelling work that indicates that there is potentially a solution for a high fusion performancedevice at relatively small major radius and low aspect ratio. In section 3, we give an overview

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PAPER EX/P3 - 36of the TE development programme; we include a brief description of the STs presently underconstruction and planned at TE. Our predictions of the performance of a candidate ST fusion
module are extended and updated in section 4. Possibilities for modular fusion are discussed
briefly in section 5. A summary is given in section 6.
2. Power and Size of Tokamak Pilot Plants and Reactors
Recent modelling with a system code based on an established physics model has shown that,when operated at reasonable fractions of the density and beta limits, tokamak pilot plants andreactors have a power gain, Qfus, that is only weakly dependent on size; mainly it depends on
Pfus, and H, where Pfus is the fusion power and H is the confinement enhancement factor
relative to empirical scalings [4]. Frequently the ITER reference scaling (IPB98y2) is usedand H is defined relative to that. When expressed in dimensionless variables this scaling has asignificant inverse dependence on the plasma beta, (β-0.9). However, dedicated experiments on
several devices in which the dependence of the confinement time on beta has been probed
directly, have shown that the confinement time is almost independent of beta; alternative betaindependent scalings have been developed, for example that by Petty [5]. These scalings arearguably more appropriate because they give consistency between the single device andmulti-device experiments. Modelling with the system code has shown that the power needed
for a given fusion gain is a factor of two to four lower with these scalings [4].
The dependence on Pfus implies that it is principally engineering and technological aspects,
such as wall and divertor loads, rather than physics considerations, that determine theminimum device size. The lower power requirement arising from the beta-independentscalings is especially advantageous. Using the system code, a wide parameter scan wasundertaken to establish possible regions of parameter space that could potentially offer highQfus with acceptable engineering parameters. In addition to the high aspect ratio, large
tokamak solution, a region of parameter space at low aspect ratio and relatively small major
radius, and hence small plasma volume, was identified (Figure 1). A candidate device

FIG. 1. Pfus as a function of R0 at constant Qfus = 30, H = 1.5 for both IPB98y2 scaling and betaindependent scalings for A = 3.2 and A = 1.8. Here Pdiv is the transported loss power that has tobe handled in the divertor after allowance for radiation losses. Details given in [4]. The
conventional large tokamak solution (left) and the potential low A solution (right) are indicated.

(ST135) with a major radius (R0) of 1.35 m, aspect ratio (A) of 1.8 and magnetic field on axis
(BT) of 3.7 T operating at 185 MW with a Qfus of 5 was suggested. The study that led to this
proposal was mainly a physics study; engineering aspects were not investigated. Some
important engineering and technological aspects are currently being developed and key results
are presented below (Section 4).

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3. Tokamak Energy Experimental Development Programme
3.1. HTS
Use of conventional low temperature superconductor (LTS) for an ST fusion device appears
impractical because thick shielding (≥ 1 m) would be needed to prevent neutrons heating thesuperconductor above 4K. With shielding of this thickness on the inner central column, thedevice would be very large. The advent of HTS, however, appears to provide a solution. Hightemperature superconductors were discovered in the late 1980’s and the 2nd generation
ReBCO (where Re=Yttrium or Gadolinium) tapes have very promising properties; inparticular, they are able to carry high currents under very high magnetic fields. Althoughsuperconductivity occurs at around 91K in zero magnetic field, far better performance isachieved when cooled to around 20 - 40 K. Thus, for constructing tokamaks, HTS has
potentially two advantages relative to LTS: an ability to carry more current at high field, andless demanding cryogenics [6].
3.2. ST25(HTS)In order to gain experience with constructing tokamaks using magnets made from HTS,Tokamak Energy constructed a small but complete tokamak. This is the world’s first
demonstration of a tokamak magnet where all the magnets are made from HTS. All coils(toroidal and poloidal) are wound from YBCO HTS tape. The 6 limb cryostat is cooled ‘cryofree’ to ~20 K using a single Sumitomo cold head seen above the vessel, thermal conduction
from the HTS tape being provided by copper strips; the 2 PF coils being cooled by He gas to
20 - 50 K. A 29 hour run was obtained in June 2015,
with a hydrogen plasma (Figure 2). The TF magnet
in ST25(HTS) uses a continuous length of YBCO
tape and, to save cost, is of low field (~ 0.1 T),
chosen to permit current drive via 2.45 GHz
microwave sources. This simple design is prone tosingle point failure at any of the several solderedjoints, and also has a high inductance, which makes
quench protection difficult. A high performance
fusion ST will need a toroidal field of 3 – 4 T, which
requires the development of HTS cables, and TE is
FIG.2. Demonstration of 29hr pulse
currently developing HTS cable and joint
in ST25(HTS) in June 2015
technologies required to build and operate a larger
device that will operate at higher field (section 3.4). A major challenge is the design of thecentre column, and this is being addressed in the design work for ST135 (below).
3.3. ST40
To date STs have operated at 1 T or below. For high fusion performance, devices operating at
3 T or above will be needed. In order to construct an ST that can operate at fields at this level
innovative engineering solutions will be needed especially for the central column. To develop
and demonstrate solutions to the key engineering aspects, TE is constructing a device (ST40)
with copper magnets that is intended to operate at fields up to 3 T. Beyond this device TE isplanning high field STs using HTS.
ST40 (Figure 3), will have a design field of BT0 = 3 T at major radius of R0 = 40 cm. Use of
copper for the TF coil (as in all existing STs, except ST25HTS at TE) has the advantages ofcombining structural strength with good conductivity (especially when cooled to liquid

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nitrogen temperature). Whereas existing STs have operated typically at 0.3 - 0.5 T, with therecent MAST, Globus-M and NSTX upgrades striving for 1 T, special design features are
employed to enable ST40 to operate at up to 3 T. Principal amongst these is the use ofConstant Tension Curve TF limbs, specially designed so that over the permitted temperaturerise (whether starting from ambient or from liquid nitrogen temperature) the expansions of thecentre post and the return limbs are matched, so that minimal movement occurs at the critical
top and bottom joints, a simple robust flexi-joint being provided to accommodate themovement. At fields of 3 T, stresses are
high; and an external support structure
based on two steel rings (shown in grey
above and below the magnet)
accommodates in-plane and out of plane
forces, such as those arising fromtolerance errors in the radial position,and the JxB twists arising from TF-PF
and TF-solenoid interactions.
An important aid to obtaining such ahigh field is the use of a minimalsolenoid. The merging compression(MC) coils (indicated in Figure 3)
FIG. 3 Engineering drawing of ST40
operated successfully in START and
MAST, and in ST40 are capable ofgenerating a hot ST plasma of up to 2
MA without use of the central solenoid, which is thus only needed to maintain the flat-top
current – assisted by the high bootstrap fractions expected, and current drive from NBI or RF.
Hence, the solenoid is considerably smaller than in MAST and NSTX and their upgrades.
This reduces JxB twisting stresses, takes less copper from the TF column thus reducing TF
resistance and TF heating, and provides a stronger TF post.
The centre post is constructed from 24 wedges, each twisted by 15 degrees over their length
thus obviating the need for a TF compensating coil. Except for the MC coils (which require afast HV bank), the TF, solenoid and PF coils are powered by ‘Supercapacitors’ such as theMaxwell 125 V, 63 F, 0.5 MJ transport module, providing a very economic power supply
from laboratory power supplies. Each unit has a limiting fault current ~ 7 kA even under deadshort conditions providing safety; an important consideration in a 100 MJ capacitor bank.
In addition to the original objective of providing a high vacuum version of the pioneering
START ST at a tenfold increase in TF, the specification has been extended: indeed it ispredicted that the MC scheme will provide 1 – 10 keV plasmas in ST40. Details of its
predicted performance are provided in a parallel paper at this conference [7]. The device iscurrently under construction and is expected to begin operation in early 2017.
3.4. Beyond ST40 (ST60)
As mentioned above, the intention is to combine the experience gained with the low field
HTS device ST25(HTS) with that obtained with the high field copper device ST40 to designand construct a high field ST using HTS for the magnets. Currently the major radius being
considered is 60 cm (ST60). The device will be used to develop physics understanding of ahigh field ST and to test HTS cable technology. It will use a design of the HTS core similar tothat outlined in section 4, which is under study for the fusion power module (ST135). A
neutron shield on the central column will probably not be included although the possibility of5PAPER EX/P3 - 36

operation with DD or DT is not excluded. For physics tests without neutrons, the device could
be operated in steady state using external current drive; in DD and DT operation, neutrons
will heat the HTS core. To avoid large cryogenic cooling costs, adiabatic short-pulse
operation is considered, whereby the HTS column is permitted to heat by 10 K before thedischarge is terminated. Initial estimates indicate that this will give minutes of operation
under DD operation; seconds if DT were to be used.
4. Conceptual Design of a Prototype Fusion Power Module: ST135
While from a physics perspective it seems that a compact fusion module may be possible(Section 2), the feasibility of such a device depends critically on there being satisfactory
engineering solutions in a few critical areas. Three important components are the central
column where it is necessary to handle the stress in this component at the same time asaccommodating the HTS TF magnet; the divertor where it is necessary to handle high powerloads; the inboard shielding that is needed to protect the HTS tape from bombardment fromhigh energy neutrons so that it has an acceptable lifetime, and also to reduce the neutron
heating to a level that can be handled with a reasonable cryogenic system. Possible solutions
for these components are under study and development within TE.
4.1. HTS Central Column Design
One possible arrangement (Figure 4) utilises several significant features of HTS tape: namely,
operation at 20 – 30 K that gives sufficient current
HTS cable (stack of HTS tapes)
carrying capability at high magnetic field, but at
Copper stabiliser
much lower cryogenic cooling cost than operation
at 4K, and the property that tape aligned parallel
Cooling channel
to the local magnetic field can carry typically afactor 4 – 5 times more current than non-aligned
20T
tape. In this simple model the individual HTS
tapes are bonded into multi-layer cables, and for
16.9T
the initial calculations we assume that the entire
structure has the strength of half-hard copper.
13.9T
This layout is chosen to provide a uniform current
10.7T
density. Towards the geometric centre of the
4.4T
7.6T
column, the magnetic field reduces and inconsequence the current carrying capacity of the
FIG. 4. Example of a
HTS tape increases. This makes it possible tomonolithic HTS centrepost.
reduce the number of tapes. The peak crushingpressure (σcc) occurs at the column axis. For this simplified model = 0.8 (/  ) MPa………………….………(1)where BT0 is in T, and major radius R0 and the HTS magnet radius Rcc are in m. For thereference ST135 design, R0 = 1.35 m, Rcc = 0.25 m, BT0 = 3.7 T, A = 1.8, and so the peak fieldat the edge of the HTS magnet is 20 T and the centre column current is 25 MA. The
calculated peak radial stress is 320 MPa. This is high, but is in the form of uniform
hydrostatic compression. There will undoubtedly be stress raisers in a real structure, but there
is a high degree of symmetry and more detailed FEA calculations indicate that these can beaccommodated.

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Expression (1) shows that forces increase as the square of the toroidal field, but reduce as thesquare of the central column radius. Hence for example, a 0.05 m addition (20%) to the HTS
core radius (accompanied by a 0.05 m decrease in shield thickness, if it is desired to maintainthe aspect ratio of 1.8, and a reduction to 16.7 T at Rcc), reduces the peak stress to 205 MPa
whilst maintaining a field of 3.7 T at R = 1.35 m.
Note that whereas it is conventional to twist superconductor cables to minimize AC losses
these will not be significant in the TF magnet of an ST power plant as this will have a slow
rise to reach a constant peak current; so with a suitable design, use can be made of thesubstantial increase in performance afforded by aligned operation, giving a substantial
reduction in cost.
4.2. Divertor Loads
High power plasmas in relatively small devices would impose very large divertor loads if
operated in single-null (SN) configuration, especially in an ST where the inner strike point is
at low radius and space to mitigate the power load by angled strike points or long divertor
legs is limited. However the use of Double-Null Divertor (DND) operation, as studied
extensively on the START and MAST STs [8], can considerably improve the loading, as the
DND configuration is very favourable for the ST concept. Firstly, the inner SOL is now(largely) isolated from the outer, and it is found that most SOL power escapes through theouter segment and so hits the outer strike points; the ratio varies widely, dependent on plasma
conditions. During ELMS the ratio can be over 20 times higher; during inter-ELM periods
when the core heating is partially retained, the ratio can fall to 4, approximately the ratio of
the inner and outer SOL areas; but the average ratio is typically taken as 10 in MAST [9].
An extensive study of divertor loads in the FNSF design [3] which is similar to ST135
estimates peak divertor loads for both the inner and outer DND strikepoints to be within 10
MW/m2, and it is instructive to compare ST135, designed to have Pfus = 200 MW and Qfus = 5,
with the FNSF design, where Pfus = 160M W and Qfus = 2. The load on the outer divertor
plates is Psol/strike area, where the strike area is proportional to Rtrg x fx x λq where Rtrg is theradius of the strike point, fx the flux expansion from the midplane to the target, and λq
represents the width of the SOL at the midplane as given by Eich scaling [10]. The heat
entering the plasma is the combination of alpha heating and auxiliary heating making a totalof 112 MW in FNSF and 80 MW in ST135, and after radiation losses due to impurity
radiation, bremsstrahlung and cyclotron this will enter the SOL. The strike points Rtrg are atabout 20% larger radius in FNSF; expansion fx should be very similar; and Eich scaling
predicts λq varies as Bpol-1.2 and Bpol is a factor 1.3 higher in FNSF and so λq is a factor 1.3
larger in FNSF. Overall, we conclude that strike areas should be very similar; and since Psol isapproximately 30% less, peak power on each outer divertor in ST135 should be ~ 7 MW/m2.
4.3. Inner Shield: Heat Deposition to Central Core and Optimisation of Shield Thickness
An extensive investigation of candidate materials for the inner shield has been carried out andtungsten carbide with water cooling has been identified as a promising configuration [11].
MCNP calculations of the attenuation due to this shield have been carried out. The attenuationof the neutron flux, and associated heat deposition in the central core, as a function of shieldthickness have been parameterised and included in the TE System Code [11]. The heat
deposited will have to be removed actively with a cryoplant and an estimate of the power
requirement is also included.

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In order to determine the optimum shield thickness several factors have to be taken into
account simultaneously. For a device of given Qfus, H factor and aspect ratio A, it is necessary
to consider each of the attenuation due to the shield, the magnetic field on the HTS tape andthe radial stress in the central column. The TE system code has been extended so that thesedifferent aspects can be taken into account simultaneously.
As an example, for a reference plasma (Qfus = 5, Pfus = 201 MW, H(IPB98y2) = 1.88, A = 1.8,
βN = 4.5), we present in Figure 5 the variation of key parameters with major radius. The extra
space in the radial build as the major radius increases is used to increase both the thickness of

FIG. 5. Heating power deposited in the superconducting core, and other key parameters, as afunction of R0. The scan has been performed with the wall loading adjusted to give a constant
HIPB98y2= 1.9, the central temperature adjusted to give 0.8 of the Greenwald density limit, andthe toroidal field adjusted to give 0.9 of the beta limit. The circles show the reference design at
1.35 m major radius. Note that Qfus, Pfus and H all remain constant as R increases.

the shield and of the HTS core in the ratio: 92% to shield, 8% to the HTS core radius Rcc whichapproximately maintains constant stress. At the reference major radius for ST135 (R0 = 1.35 m),
the HTS core has radius 0.25m, gaps total 0.05m, the shield thickness is 0.3 m, the field on
the conductor is 20 T, the neutron heating to the central column is 100 kW, and the wall load
is 2 MW/m2. To handle this level of neutron heating we estimate that a cryogenic plant of 5
MW wallplug power would be needed. It is clear from the figure that as the shield thickness
increases the heating of the central column reduces rapidly. Using expression (1), the stress inthe central column is 320 MPa. As discussed in section 4.1, if the scan was repeated with a 5
cm exchange increasing Rcc at the expense of shield thickness, the stress would be reduced to
205 MPa but Pheat and Pcry would increase. From the comparison with FNSF (section 4.2) the
peak divertor load would be ~ 7 MW/m2. For all values of R0, q*eng (defined as 5(1+κ2)/2
a2B/(RI) ) is ≥ 2.8, the value recommended for avoidance of disruptions [3].
A key aspect not yet included is any change in tape performance due to irradiation by
neutrons. Inevitably the performance will degrade but information on the extent of thedegradation is limited. Dedicated R&D is needed in this area.

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5. A Modular Power Plant
If a relatively small fusion module is feasible, then a possible alternative supply of fusion
power based on a modular concept may be available. The economics and operational
advantages of this concept, utilizing perhaps 11 small 100 MW units, (10 working and 1
undergoing maintenance) have been outlined [12]. The advantages include improved
availability; cyclic maintenance; the need for only a relatively small hot cell; a sharing ofstart-up facilities and energy conversion; the possibility of pulsed operation to provide loads
variable in time and the economics of mass-production. STs can exhibit the combination ofhigh bootstrap fraction and high beta - important both for maximizing power gain and in
obtaining/maintaining the plasma current, especially in the absence of a central solenoid. In
this latter respect, recent predictions that RF techniques can provide full plasma current
initiation and ramp-up [13] are encouraging; it is expected that these techniques will perform
well in a high field (3 - 4 T) ST as then the densities can be sufficiently high.
6. Summary
The Tokamak Energy programme aimed at developing the spherical tokamak as a future
power source is outlined. Areas that have a high leverage on the feasibility of this approach
have been identified and are under study in current R&D. Two such areas are the energy
confinement scaling at high field (3 – 4 T), and the impact of fusion neutron irradiation on theproperties of HTS rare earth tape at 20 - 30 K. Both are under investigation and the data
should be available in the near future. Favourable results could lead to economic fusion basedon modular high gain STs of small size; less favourable results could lead to larger but still
economic ST fusion power plants of around 1.5 - 2 m major radius. In either case, the small
scale of the fusion modules should lead to rapid development and make possible theresolution of the remaining key outstanding physics and technology steps that are needed forthe realisation of fusion power.
7. References
PENG, Y-K.M., STRICKLER, D.J. Nuclear Fusion 26 (1986) 769
SYKES, A. et al., Nuclear Fusion. 39 (1999) 1271.
MENARD, J.E., BROWN, T., EL-GUEBALY, L., et al ‘Fusion nuclear sciencefacilities and pilot plants based on the spherical tokamak’ Nuclear Fusion, 56, Number
10, 2016[4] COSTLEY, A.E., HUGILL, J. and P BUXTON, P. Nuclear Fusion 55 (2015) 033001[5] PETTY, C.C. Phys. Plasmas 15 (2008) 080501[6] SYKES, A. et al, ‘Recent advances on the Spherical Tokamak route to Fusion Power’ ,IEEE Transactions on Plasma Science 42 , (2014) 482 – 488[7] GRYAZNEVICH, M.P. et al, ‘Overview and status of construction of ST40’, submitted
to this conference[8] COUNSELL, G.F. et al, ‘Boundary plasma and divertor phenomena in MAST’, PP&CF
44 (2002) 827-843[9] FISHPOOL, G. et al, P1-11 ‘MAST-Upgrade Divertor Facility and Assessing…’
arxiv.org/pdf/1306.6774.pdf; 2013[10] EICH,T. et al., Nucl. Fusion 53 (2013) 093031[11] WINDSOR, C. G., MORGAN, J. G. & BUXTON, P.F. Nucl. Fusion 55, 023014(10pp)(2015).
[12] GRYAZNEVICH, M.P. and CHUYANOV, V.A., J.Phys Conf. Ser. 591 (2015) 012005.
[13] RAMAN, R. & SHEVCHENKO, V., Plasma Phys. & Contr. Fus. 56 (2014) 103001.
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