Spherical tokamaks (STs) have features that make them a potentially attractive option for fusion power production compared to conventional tokamaks (CTs) including operation at high beta and high self-driven ‘bootstrap’ current. The thermal energy confinement time ( τ Ε ) also typically has a stronger dependence on toroidal magnetic field and a weaker dependence on plasma current, but so far it has not been established how this difference impacts performance under reactor conditions. This aspect is explored in this paper. Using empirical data from NSTX and MAST, and from multiple CTs, we investigate analytically and by using established fusion codes the potential fusion performance, characterised by the fusion triple product, nTτ Ε , and fusion power gain, Q fus , where n and T are the density and temperature respectively. We find that for similar values of field and fusion power, but smaller volume, STs can have nTτ Ε up to a factor of three higher and Q fus an order of magnitude higher than CTs. We identify the origin of this enhanced performance and outline a measurement to advance this finding. Potentially our results open an alternative and faster route to fusion power based on relatively small, low power STs.

Despite several decades of dedicated R&D, fusion, a potentially world-changing energy source, remains decades away from commercialisation. The majority of development thus far has been via publicly-funded programmes led by government laboratories focused on scientific research and in which commercialisation strategy and innovation play a minor role. Generally, such programmes follow a linear model of innovation in which commercial aspects are not considered until later in development. In consequence and without intention, devices not well-suited for commercial application are being pursued. In recent years, however, privately funded fusion start-ups have emerged with the goal of accelerating the commercialisation of fusion. Fusion start-ups are, by necessity, operating on a fundamentally different model of innovation: agile innovation, whereby technology is developed flexibly and iteratively towards an explicit commercial goal. Technology Roadmapping is a method that has been effective for supporting agile innovation but thus far has had limited application to mission-led hardware development. We characterise the key features of the fusion innovation approach and create a novel Technology Roadmapping process for fusion start-ups, which is developed via a case study with Tokamak Energy Ltd. The main elements of the developed process, the resulting Technology Roadmap, and its impact are presented.

The question of size of a tokamak fusion reactor is central to current fusion research especially with the large device, ITER, under construction and even larger DEMO reactors under initial engineering design. In this paper, the question of size is addressed initially from a physics perspective. It is shown that in addition to size, field and plasma shape are important too, and shape can be a significant factor. For a spherical tokamak (ST), the elongated shape leads to significant reductions in major radius and/or field for comparable fusion performance. Further, it is shown that when the density limit is taken into account, the relationship between fusion power and fusion gain is almost independent of size, implying that relatively small, high performance reactors should be possible. In order to realize a small, high performance fusion module based on the ST, feasible solutions to several key technical challenges must be developed. These are identified and possible design solutions outlined. The results of the physics, technical and engineering studies are integrated using the Tokamak Energy system code, and the results of a scoping study are reviewed. The results indicate that a relatively small ST using high temperature superconductor magnets should be feasible and may provide an alternative, possibly faster, ‘small modular’ route to fusion power. This article is part of a discussion meeting issue ‘Fusion energy using tokamaks: can development be accelerated?’.

Experiments on NSTX and MAST have shown the thermal energy confinement time in spherical tokmaks (STs), tau(E,th), to have a stronger toroidal field and weaker plasma current dependence than in conventional large aspect ratio tokamaks. These scalings were derived for single machines both of which are similarly sized, consequently the NSTX and MAST scaling laws do not include a size dependence, and so cannot be used to extrapolate the performance of future STs. Using physics-based dimensional arguments we extend the NSTX scaling to include a size scaling. We also resolve a colinearity problem specific to NSTX data by assuming core transport is gyro-Bohm like. The resulting scaling has approximately zero beta dependence, a typical collisionality dependence, and a relatively weak safety factor dependence: B tau(E,th) proportional to rho(-3)(*)nu(-0.53)(*) * beta(-0.17) q(-0.35). With the exception of the safety factor, all exponents are consistent with recent experiments in large aspect ratio tokamaks. This apparent difference between STs and large aspect ratio tokamaks is consistent with MAST and NSTX results. We have considered the implications of the scaling for pilot plants and reactors and find it may be possible to develop more compact reactors based on the ST approach.

The fast ion and alpha-particle diagnostic is based on collective Thomson scattering of high power 140 GHz radiation.The main aim of the system will be the determination of the spatially resolved velocity distribution of fast ion populationsin JET plasmas, by measuring the frequency spectrum of scattered radiation. A spatial profile of the velocity distribution,with resolution <10 cm, can be determined by scanning the scattering volume (the overlap of the antenna patterns of thelaunch and receive antennas) over the plasma, using steerable launch and receive mirrors. The diagnostic will be used tomeasure fast ion populations produced by the heating systems, and will be particularly important in the tritium phase ofJET when, by a measurement of the velocity distribution of alpha particles created by fusion reactions, the efficiency ofalpha particle heating can be established. It is expected that their velocity distribution can be determined between 0.5MeV (limited because at low frequency shift the spectrum is dominated by the thermal ion feature) and their birth energy(3.5 MeV) with a typical time resolution of 0.1 s. The principal components of the diagnostic are shown schematically infigure 1. They are: a high power long pulse gyrotron and associated equipment, a heterodyne receiver system andtransmission lines. The RF transmission and the receiver system are described in accompanying papers1,2. The physicsprinciples of the diagnostic are described elsewheree.g. 3'4.

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.

Numerical studies have been made to improve the performance of the central column of a superconducting spherical tokamak fusion pilot plant. The assumed neutron shield includes concentric layers of tungsten carbide and water. The relative thickness of the water layers was varied and a minimum power deposition was found at about 17% of water. It was found advantageous to have an approximately 1.7 times thicker water layer next to the core and a similarly thinner layer next to the plasma. The use of tungsten boride instead of tungsten carbide was shown to make an improvement especially if placed close to the central superconducting core, the inner layer alone reducing the power deposition by 29%. Engineering features such as a central steel tie-bar, an insulating thermal vacuum gap, a wall gap next to the plasma and knowledge of the vertical energy distribution are essential to a successful design and their effects on the power deposition are shown in an appendix. The results have been fitted to model distributions and incorporated into the Tokamak Energy System Code, which can then give predictions of the power deposition as a function of other parameters such as the plasma major radius and the maximum magnetic field permitted on the superconductors.

In this comment, we discuss the arguments raised in two recent papers (Costley 2016 Nucl. Fusion 56 066003, Costley et al 2015 Nucl. Fusion 55 033001) on the claimed size independence of fusion power, triple product and fusion gain in tokamak reactors, and we show that all these three quantities actually do depend on the size of the tokamak, when distinguishing between independent input parameters (design parameters) and output quantities, and when taking into account technological limitations.

In reply to Frank Chen's letter "Big fusion, little China" (August 2016) about the size of fusion reactors.

The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτE, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τE are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτE becomes almost independent of size; rather it depends mainly on the fusion power, Pfus. In consequence, the fusion power gain, Qfus, a parameter closely linked to nTτE is also independent of size. Hence, Pfus and Qfus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτE is inversely dependent on the normalised beta, βN; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

In reply to the news story "Smaller fusion reactors could deliver big gains" (March p12), which refers to a study led by Alan Costley of the UK firm Tokamak Energy.

The realization of power from Magnetic Confinement Fusion is presently following a plan that will span several decades. The mainstream route is via ITER, combined with, or followed by, materials development on a Fusion Neutron Science Facility, then a DEMO reactor, which, as presently visualised, is considerably larger than ITER. We consider smaller-scale alternatives and developments which may make more rapid progress towards the much needed goal of economic, safe, clean fusion power, but still based on the tokamak.

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher B-T and I-p can result in a high gain Q(fus) similar to 12, P-fus similar to 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) > 1 as a target.

It is generally accepted that the route to fusion power involves large devices of ITER scale or larger. However, we show, contrary to expectations, that for steady state tokamaks operating at fixed fractions of the density and beta limits, the fusion gain, Qfus, depends mainly on the absolute level of the fusion power and the energy confinement, and only weakly on the device size. Our investigations are carried out using a system code and also by analytical means. Further, we show that for the two qualitatively different global scalings that have been developed to fit the data contained in the ITER ELMy H-mode database, i.e. the normally used beta-dependent IPB98y2 scaling and the alternative beta-independent scalings, the power needed for high fusion performance differs substantially, typically by factors of three to four. Taken together, these two findings imply that lower power, smaller, and hence potentially lower cost, pilot plants and reactors than currently envisaged may be possible. The main parameters of a candidate low power (∼180 MW), high Qfus (∼5), relatively small (∼1.35 m major radius) device are given.

Stambaugh developed the Peng-Hicks concept of a fusion reactor based on a solid copper center-post spherical tokamak (ST). Using the promising results from the START experiment, they produced a vision for a path to fusion power. This path had two elements such as the ability to produce high fusion gain from an ST and of equal importance, the ability to demonstrate this in a small (and therefore relatively low cost) pilot plant device. In this paper, we review various attempts to pursue this vision, and try to elucidate the reason why success has not yet been achieved. However, we show that the advent of high temperature superconductors may overcome some of the problems, and we suggest a revised version of the small, low entry cost route to fusion power.

Diagnostic instruments are needed to provide measurements for machine protection (magnet systems, vacuum systems, heating and cooling systems, safety systems, etc.); for plasma control; and for plasma performance evaluation. Plasma diagnostic systems include electrostatic probes, magnetic probes, measurement of particles emitted from the plasma, spectroscopy of waves emitted by the plasma, probing the plasma with injected particle beams, and probing the plasma with waves, such as microwaves and laser beams. For ITER, the next major step in magnetic fusion, diagnostics must be hardened to survive in a hot, high-radiation environment, and designed to minimize neutron streaming through ducts. Many adverse effects must be mitigated, such as radiation-induced spurious signals, erosion of windows and mirrors, and deposition of films on them. This is new territory for diagnostics and many challenges have to be overcome.

A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS).With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.

In ITER, mirrors will be used as plasma-viewing elements in all optical and laser diagnostics. In the harsh environment mirror performance will degrade hampering the operation of associated diagnostics. The most adverse effect on mirror performance is caused by the deposition of impurities. It is expected that the most challenging situation will occur in the divertor. With the envisaged changes to all-metal plasma-facing components (PFCs) in ITER, an assessment of mirror performance in an existing divertor tokamak with allmetal PFCs is urgently needed. Molybdenum and copper mirrors were exposed for nearly nine months in ASDEX Upgrade which has alltungsten PFCs. Mirrors were located at the inner wall, under the dome and in the pump-duct. During exposure, the mirrors were heated to temperature in the range 145C-165C. This was made to approach the expected level of heating due to absorption of neutrons and gammas on mirrors in ITER divertor. After exposure, degradation of the reflectivity was detected on all mirrors. The highest reflectivity drop was measured on mirrors under the dome facing the outer strike point, reaching -55% at 500 nm. The least degradation was detected on mirrors in the pump duct, where the reflectivity was preserved in the range 500-2500 nm and the largest decrease was about -8% at 250 nm. On all contaminated mirrors carbon fraction did not exceed 50 at.% while the major contaminants were metals and oxygen. The degradation of exposed mirrors underlines the necessity for urgent R&D on deposition mitigation and in-situ mirror cleaning in ITER.

In ITER, mirrors will be used as plasma-viewing elements in all optical and laser diagnostics. In the harsh environment, mirror performance will degrade hampering the operation of associated diagnostics. The most adverse effect on mirror performance is caused by the deposition of impurities. It is expected that the most challenging situation will occur in the divertor. With the envisaged changes to all-metal plasma-facing components (PFCs) in ITER, an assessment of mirror performance in an existing divertor tokamak with all-metal PFCs is urgently needed.Molybdenum and copper mirrors were exposed for nearly nine months in ASDEX Upgrade which has all-tungsten PFCs. Mirrors were located at the inner wall, under the dome and in the pump duct. During exposure, the mirrors were heated to temperature in the range 145-165 degrees C. This was made to approach the expected level of heating due to absorption of neutrons and gammas on mirrors in the ITER divertor. After exposure, degradation of the reflectivity was detected on all mirrors. The highest reflectivity drop was measured on mirrors under the dome facing the outer strike point, reaching -55% at 500 nm. The least degradation was detected on mirrors in the pump duct, where the reflectivity was preserved in the range 500-2500 nm and the largest decrease was about -8% at 250 nm. On all contaminated mirrors carbon fraction did not exceed 50 at% while the major contaminants were metals and oxygen. The degradation of exposed mirrors underlines the necessity for urgent R&D on deposition mitigation and in situ mirror cleaning in ITER.