Fusion Energy based on the Spherical Tokamak

A. Sykes¹, A.E. Costley¹, C.G, Windsor¹, O. Asunta¹, G. Brittles¹, P. Buxton¹, V. Chuyanov, J.W. Connor¹, M.P. Gryaznevich¹, B. Huang¹, J. Hugill¹, A. Kukushkin2,3, D. Kingham¹, A.V. S. McNamara¹, J.G. Morgan⁴, P. Noonan¹, J.S.H. Ross1¹, V. Shevchenko¹, R. Slade¹and G. Smith^1,5
1Tokamak Energy, D5, Culham Science Centre, Abingdon, OX14 3DB
²NRC "Kurchatov Institute", Kurchatov Sq.1, 123182 Moscow, Russia
³MRNU MEPhI, Kashirskoje Ave. 31,115409 Moscow, Russia
⁴Culham Electromagnetics, D5, Culham Science Centre, Abingdon, OX14 3DB
⁵Department of Materials, University of Oxford, 16 Parks Road, Oxford OX1 3PH UK
E-mail: alan.sykes@tokamakenergy.co.uk

15 December 2016, revised 19 July 2017
Accepted for publication 14 September 2017

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 the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

Keywords: fusion energy, spherical tokamak, high temperature superconductor

1. Introduction

Since the mid -1980s the spherical tokamak (ST) has been recognized as an important device for fusion research [1-4]. 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 [2], can be run at higher elongation [3, 4], and possibly exhibit higher confinement [4], although more data are needed at higher field and lower collisionality to determine this important aspect. Early attempts to design reactors based on STs did not produce convincing designs, and until 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 and superconductor technology have changed the situation, and relatively small STs operating at high fusion gain are now considered possible. The key physics step is the realization that the power and the device size 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 withstand relatively high magnetic fields: both of these properties are beneficial in the design of magnets for fusion devices especially for STs where space is limited in the central column. Sorbom et al [5] have considered the application of HTS to tokamaks of conventional aspect ratio and produced a design for ARC, a fusion power plant slightly greater in size than JET and at considerably higher field. In this paper, we describe the tokamak energy (TE) programme to develop an alternative route to fusion power based on STs constructed using HTS magnets, and the modelling and concept work 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 is underway addressing these aspects, including the construction and operation of a series of STs. 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 performance device at relatively small major radius and low aspect ratio. In section 3, we give an overview of the TE development programme; we include a brief description of the STs operated, presently under construction 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

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 and reactors have a power gain, Q_fus , that is only weakly dependent on size; mainly it depends on P_fus , and H, where P_fus is the fusion power and H is the confinement enhancement factor relative to empirical scalings [6]. Frequently the ITER reference scaling (IPB98y2) is used and H is defined relative to that. When expressed in dimensionless variables this scaling has a significant 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 beta-independent scalings have been developed, for example that by Petty [7]. These scalings are arguably more appropriate because they give consistency between single device and multi-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 (figure 1) [6].

The dependence on P_fus implies that it is principally engineering and technological aspects, such as wall and divertor loads, rather than physics considerations, that determine the minimum device size. The lower power requirement arising from the beta-independent scalings is especially advantageous. Using the system code, a wide parameter scan was undertaken to establish possible regions of parameter space that could potentially offer high Q_fus 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, has been identified (figure 1). The physics advantages (such as high beta) of low aspect ratio potentially enable a compact ST module to achieve a high fusion gain at a modest toroidal field (TF) of around 4 T, whereas a compact conventional aspect ratio tokamak requires a very high field on axis ~12 T to achieve high fusion gain as evidenced by Ignitor [15]. A candidate device (ST135) with a major radius (R₀) of 1.35 m, aspect ratio (A) of 1.8 and magnetic field on axis (B_T0) of 3.7 T operating at P_fus = 185 MW with a Q_fus 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 in this paper (section 4).

3. TE 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 the superconductor to above 4 K. With shielding of this thickness on the inner central column, the device would be very large. The advent of HTS, however, potentially provides a solution. HTSs were discovered in the late 1980s and the 2nd generation ReBCO (where Re = Yttrium or Gadolinium) tapes have very promising properties; in particular, they are able to carry high currents under very high magnetic fields. Although superconductivity occurs at around 91 K in zero magnetic field, far better performance is achieved 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, and less demanding cryogenics [8].

Figure 1. P_fus as a function of R₀ at constant Q_fus = 30, H = 1.5 for both IPB98y2 scaling and beta independent scalings for A = 3.2 and A= 1.8. The values of some key engineering parameters are given in the text in the figure including the field at the conductor on the inboard side, B_cond. P_div is the transported loss power that has to be handled in the divertor after allowance for radiation losses. Details are given in [6]. The conventional large tokamak solution (left) and the potential low A solutions (right) are indicated. The circled areas show that, with the beta-independent scaling, the wall loads and divertor loads for a relatively small (R = 1.4 m) low aspect ratio (1.8) device would be in the range of 3.5 MW m^-2 and 45 MW m^-1 respectively, for the same Q_fus . This is challenging (although proposals to reduce the divertor load in an ST are described in section 4.2) but in the region of those likely to have to be dealt with in much larger and more powerful devices. The value of B_cond is high in the case of the low A approach but potentially achievable using magnets made with HTS (sections 3 and 4).

3.2. ST25(HTS)

To gain experience with constructing tokamaks using magnets made from HTS, TE constructed a small but complete tokamak (figure 2). This provided 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 TF cryostat is cooled 'cryo-free' 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 two poloidal field (PF) coils being cooled by He gas to 20-50 K. A 29 h run was obtained in June 2015, with an RF discharge in hydrogen (figure 2). The TF magnet in ST25(HTS) used a continuous length of 12 mm wide YBCO tape of 48 turns in each of 6 limbs which when operated at 400 A would provide a TF at R = 0.25 m of ~0.1 T, chosen to permit current drive (CD) via 2.45 GHz microwave sources. This simple design is prone to single point failure (particularly at any of the several soldered joints), was not designed to tolerate quenches, and was operated considerably below the critical current which is ~1 kA at 20 K and in the low self-field which is <1 T at the inner TF limb. A high performance fusion ST will need a TF of 3-4 T, which requires the development of high current HTS cables. TE is currently developing HTS cable, joint and quench management technologies required to build and operate a larger device that will operate at higher field and with very large stored energy (section 3.4). A major challenge is the design of the central column, and this is being addressed in the design work for ST135 (section 4.1).

3.3. ST40

To date STs have operated at TFs of less than 1 T. For high fusion performance, devices operating at 3 T or above will be needed. 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 is planning high field STs using HTS. ST40 (figure 3), will have a design field of B_T0 = 3 T at major radius of R₀ = 0.4 m, and a centre-rod current of 6 MA. Use of copper for the TF coil (as in all existing STs, except ST25HTS at TE) has the advantages of combining structural strength with good conductivity (especially when cooled to liquid nitrogen temperature). Whereas existing STs have operated typically at 0.3-0.5 T, with the recent MAST, Globus-M and NSTX upgrades striving for 1 T, innovative design features are employed to enable ST40 to operate at up to 3 T. Principal amongst these is the use of Constant Tension Curve TF limbs, specially designed so that over the permitted temperature rise (whether starting from ambient or from liquid nitrogen temperature) the expansions of the centre 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 the movement. 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 from tolerance errors in the radial position, and the JxB twists arising from TF-PF and TF-solenoid interactions. The ST40 mechanical design was analysed extensively by a series of electromagnetic analyses using Opera [10], which simulated the forces expected in operational scenarios, including Vertical Displacement Events. These forces were then used in mechanical Finite Element Analyses of major components, using Ansys [11]. For example the central column of the TF magnet, formed from 24 twisted wedge-shaped conductors, exhibits highest stress at the inner edge. For the maximum wedge current of 0.25 MA required to produce a field of 3 T at plasma major radius of 0.4 m, this stress is ~100 MPa in the copper. The copper is half hard, with a yield stress around 180 MPa. Comparing this to the Von Mises yield criterion gives a factor of safety on yield of 1.8.

An important aid to obtaining such a high field is the use in ST40 of a minimal solenoid, made possible by the merging-compression (MC) process for plasma start-up. This should produce hot plasmas with currents of up to 2 MA without use of the central solenoid, which is only needed to maintain the flat-top current-assisted by the high bootstrap fractions expected, and CD from NBI or RF. Hence, the solenoid is considerably smaller than in MAST and NSTX and their upgrades. This reduces JxB twisting stresses, allows more copper for the TF column which reduces TF resistance and heating, and provides a stronger TF post.

Figure 2. Demonstration of 29 h RF discharge in ST25(HTS) in June 2015. The 6-limb TF magnet is cooled to ~20 K by the Sumitomo cold head shown above the magnet.

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. The TF, solenoid and PF coils are powered by 'Supercapacitors' such as the Maxwell 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 dead short conditions providing safety; an important consideration in a 100 MJ capacitor bank. The plasma pulse length is limited by the temperature rise in the centre post; initial operations with a water-cooled TF magnet will provide a TF of 1-2 T at R₀ = 0.4 m with a flat top of 1-3 s; operation with liquid nitrogen cooling considerably reduces resistance and hence heating and enables longer pulses, and should permit operation at up to 3 T with a flat top of ~1 s.

The MC coils (indicated in figure 3) operated successfully in START and MAST, and extrapolation to ST40 is discussed in [12]. The MC process involves the formation of plasma rings around each of the MC coils shown in figure 3 by rapid discharge of high voltage capacitor banks. These plasma rings attract each other and merge on the midplane, followed by an adiabatic compression of the plasma to the desired major radius of ~0.4 m. It is shown that plasma current immediately after merging increases with TF and linearly with MC coil current. In ST40 the TF and MC coil current are increased over those in MAST by factors of up to 6 and 2 respectively. Extrapolation indicates ST40 should have plasma current after merging of around 1 MA; the subsequent adiabatic compression phase halves the radius in ST40 and should approximately double the plasma current, assisted by the significant reduction in inductance of the plasma ring as it takes up the highly shaped ST form. The MC scheme will be operated at the highest performance permissible, to produce the highest possible plasma currents and plasma temperatures. The final design features MC coil currents of 600 kAt in each coil, produced by a 11 kV, 28 mF capacitor bank, with a downswing time of ~10 ms which induces the plasma rings, and uses very slender support legs (to minimise interference with the plasma rings which have changing helicity). The MC coil mounting structure was analysed in Ansys, and a prototype was tested to 10 000 cycles at the design load of 73 kN, and then finally pulled to destruction. Final failure occurred at approximately 3 times the design load.

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 is expected that the MC scheme will provide up to 10 keV plasmas in ST40, the plasma being heated by the rapid conversion of magnetic field energy into plasma kinetic energy during the merging. Full details of its predicted performance, and of the expected evolution of electron and ion temperature profiles are provided in [12], based on extensive studies on both MAST and Japanese STs in collaboration with Y Ono and his team [13]. The ST40 device is currently under construction and is expected to begin operation in 2017.

3.4. Future development programme

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 design and construct high field STs using HTS for the magnets. The objectives will be to develop physics understanding of a high field ST, to test HTS cable technology, and to establish HTS performance during DT fusion conditions. ST40 should provide valuable information to determine energy confinement scaling in a high-field ST. TE is designing a high field HTS magnet, using cable technology similar to that described in section 4, to establish the engineering viability. Research is advancing rapidly on these topics, both in-house and worldwide, and the precise DT fusion experiments are still under consideration. As shown in section 4.3 these can range in size from small short pulse research devices, to steady state devices of major radius ~2 m.

Figure 3. Left: engineering drawing of ST40, showing the steel support rings above and below the vessel, and the merging-compression coils which provide an initial high current, hot plasma without need of the central solenoid. Right: model of the vacuum vessel and components of ST40. KINX simulations show [9] that growth rates of the highly elongated ( κ ~ 2.6) plasma shown can be limited to e-fold times of ~20 ms by the passive plates (indicated), and can be stabilized by internal active feedback coils.

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 as accommodating the HTS TF magnet; the divertor where it is necessary to handle high power loads; the inboard shielding which is needed to protect the HTS tape from bombardment from high 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, and are outlined in the following sections.

4.1. HTS central column design

One possible arrangement (figure 4) utilises two significant features of HTS tape: namely, operation at 20-30 K that gives sufficient current carrying capability at high magnetic field, but at much lower cryogenic cooling cost than operation at 4 K, and the property that tape aligned parallel to the local magnetic field can carry several times more current than nonaligned tape. In this simple model the individual HTS tapes are bonded into multi-layer cables, and for the initial calculations we assume that the entire structure has the strength of half-hard copper.

Towards the geometric centre of the column, the magnetic field reduces and in consequence the current carrying capacity of the HTS tape increases. This makes it possible to reduce the number of tapes. For this simplified design, in which the HTS cables are arranged to produce a uniform current density over the central HTS magnet, we can derive a simple expression for the peak stress which is at r = 0, as follows.

Current density in the centre rod magnet is J_cc = I_cc/(π R_cc²) where I_cc(MA) is the total centre rod current, and R_cc(m) is the radius of the magnet. Since we are assuming constant current density in the central column, the TF in Tesla at any radius r(m) within it is B(r) = 0.2 r I_cc/R_cc². If we neglect hoop stress and integrate the J x B force from r to R_cc we can obtain the inward force at any radius within the central column. We find that the peak compressive stress (σ_cc) occurs at the column axis, and is