Fusion energy using tokamaks: can development be accelerated?

The tokamak may be the most successful type of device in fusion research but it is far from the only one. A few other parallel strands have been pursued for decades in government labs, but recently, in part because of slow progress in tokamak development, a new breed of fusion start-up companies has emerged to seek a faster-cheaper-better route to fusion energy. Stellarators, which look superficially like tokamaks but contain plasma differently, are currently championed by Germany’s Wendelstein 7-X which recently began operation. Inertial confinement fusion uses lasers to intensely compress small capsules of fuel and spark a burst of fusion. The leader in this field, the US National Ignition Facility, is currently struggling to get the spark to light but variants such as direct-drive and fast-ignition are waiting in the wings. Magneto-inertial fusion, which uses magnetic confinement followed by explosive compression, is being pursued by US national labs Los Alamos and Sandia, as well as some fusion start-ups. Others are banking on self-confining balls of plasma known as spheromaks or field-reversed configurations, or on electrostatic confinement in a device known as a Polywell. Dan Clery hopes to show there are more ways than one to skin a fusion cat.

The European roadmap to the realisation of fusion electricity breaks the quest into eight missions. For each mission, it reviews the current status of research, identifies open issues, and proposes a research and development programme.

ITER is the key facility of the roadmap as it is expected to achieve most of the important milestones on the path to fusion power. The Fusion Roadmap is tightly connected to the ITER schedule and the vast majority of fusion resources in Horizon 2020 are dedicated to ITER and its accompanying experiments. Parallel to the ITER exploitation in the 2030s, the construction of the demonstration power plant DEMO needs to be prepared. DEMO will for the first time supply fusion electricity to the grid. The design, construction and operation of DEMO require full involvement of industry to ensure that, after a successful DEMO operation, industry can take responsibility for commercial fusion power.

This is Professor Janeschitz’s personal vision and outlook towards a fusion reactor based on his extensive experience from being part of the ITER design and now construction as well as leading the largest fusion technology program worldwide (KIT) for seven years. In particular he will to discuss how a fusion reactor can be economically viable without employing too advanced (SCFI) physics and technology. It certainly will be a pulsed machine (~ 20000 sec pulses) with thermal energy storage (turbine is steady state). He will also discuss the optimum machine size and toroidal field for such a machine and why he thinks that high field and smaller plasmas may not necessarily make a fusion reactor more competitive. 

When one extrapolates from today’s knowledge on ITER construction, even considering that ITER can be built much cheaper, it is clear that a fusion power plant will cost more than 10 or more likely more than 15 billion Euro / Dollar (the first of a kind even ~ 30 billion). Therefore in order to have an economically attractive fusion reactor it needs to produce a large amount of power (in the order of 2.5 GW electric). The possible size (R~10 m) and reasonably conservative physics basis of such a machine will be briefly described in the presentation. If we are successful in achieving advanced physics in a burning plasma, eg in ITER, then we can make the machine slightly smaller, but the principle arguments for a large machine will not change significantly. 

Key technologies and their status will be discussed with particular emphasis on a realistic blanket and divertor design and the size and issues of a T-plant for such a machine as well as the challenges which have to be overcome beyond what is needed for ITER.

At the end a simple economic consideration will be discussed to show that a large machine could be economically acceptable even in today’s environment, in particular in competition to renewables. 

The talk will outline the key challenges which must be addressed to deliver fusion power and overview the facilities which UKAEA have now, and plan for the future, to address these challenges. UKAEA is helping UK industry to win key contracts for delivery of systems and components to ITER, and the talk will also explore how this connection to industry is vital for delivering fusion power.

The technical challenges to fusion power are wide and varied, from advanced engineering and manufacturing methods, through novel materials in extreme environments to the swirling, writhing plasma we are seeking to contain. This talk will focus on just one area, that is rich in physics and sits at the heart of any fusion reactor – the plasma. In a tokamak, the plasma is confined in a toroidal geometry by magnetic field. Professor Wilson will explore how turbulence degrades the confinement, and is the main driver for pushing up the size of fusion reactors. However, there are circumstances when the turbulence is suppressed, resulting in regions of steep pressure gradient in the plasma. While good for confinement, the associated free energy drives filamentary plasma eruptions, like miniature solar flares, which on a device like ITER would cause excessive damage. This has led to methods to control these eruptions, which we shall review. The heat and particles that leak from the magnetic confinement system are handled in a region that is remote from the plasma core, called the divertor. Design of the divertor is one of the biggest challenges facing fusion, driving innovative solutions such as the 'Super-X' divertor on the UK’s MAST-U tokamak which will begin operation during 2018.

The construction of JT-60SA by Japan and Europe is progressing on schedule towards its first plasma in September 2020. As of February 2018, more than 90% of the component manufacture has been completed and closure of the vacuum vessel is close to be accomplished. The objectives of the JT-60SA project are i) contribution to the success of ITER, ii) complementation of ITER for DEMO in all major areas of plasma development, and iii) fostering the next generation who are leading ITER and DEMO. JT-60SA is a highly-shaped large superconducting tokamak (maximum plasma current 5.5 MA, major radius ~3 m, aspect ratio ~2.6, elongation ~1.8, triangularity ~ 0.4) with a variety of plasma control actuators including high power heating (41 MWx100 s). This device is capable of confining break-even-equivalent class deuterium plasmas lasting for a duration (typically 100 s) longer than the time scales characterizing the key plasma processes in the ITER and DEMO-relevant plasma regimes. JT-60SA also pursues fully non-inductive steady-state operations with high plasma pressure (beta-value) exceeding the no-wall ideal MHD stability limit. The JT-60SA Research Plan ver.3.3 has been documented by ~ 400 scientists in Japan and the EU. Nationally coordinated research is being carried out in Japanese universities using various specialized 8 Spherical Tokamak (ST) devices towards realization of a component test facility and a power reactor. Research topics pursued on these devices include plasma current start-up using RF waves and coaxial helicity injection etc, advanced fuelling using compact toroid injection, ultra-high-beta plasma formation using magnetic reconnection heating by plasma merging. Including JT-60SA, tokamak research in Japan systematically studies highly-shaped high-beta plasmas with non-inductive current drive.

Studies for conventional tokamak power plants usually end up with a major radius R0 of the order 8-9 m, ie significantly beyond the largest existing tokamak, JET (R0 = 3m), and also in excess of the size of ITER (R0 = 6.2 m). This can be understood from a simple 0-D scaling [1] by the need to operate in a state of large power amplification Q=Pfus/PAUX > 30, where Pfus is the fusion power and PAUX the auxiliary heating power to compensate for residual plasma energy loss. However, these considerations assume a technical limit of the confining magnetic field B in line with the limits of the ITER design. Recent advances in High Temperature Superconducting Coil technology have led to proposals based on higher B, leading to more compact devices, i.e. smaller R0 [2], [3].

In this contribution, Professor Zohm argues that the definition of tokamak ‘size’ should include the magnetic field to remove the ambiguity in the discussion of ‘size dependence’ of the performance of fusion power plants [3], [4]. He also analyses the possibilities that higher B would offer, using an extension of the 0-D model used in [1]. Different routes of taking advantage of higher field are discussed. It is shown that, one also has to consider consistently the assumptions about plasma performance, such as confinement quality, operational limits or exhaust schemes. Finally, Professor Zohm discusses some of the significant implications for future R&D needed to make higher magnetic field in reactor-grade devices a reality.  

[1] H. Zohm, Fusion Sci. Technology 58 (2010) 613.
[2] B. Sorbom et al., Fus. Eng. Design 100 (2015) 378.
[3] A. Costley et al., Nucl. Fusion 56 (2016) 066003.
[4] W. Biel et al., Nucl. Fusion 57 (2017) 038001.

Research in the 1970s and 80s by Sykes, Peng, Jassby and others showed the theoretical advantage of the spherical tokamak (ST) shape. Experiments on START and MAST at Culham throughout the 1990s and 2000s, alongside other international STs like NSTX at the Princeton Plasma Physics Laboratory, confirmed their increased efficiency (namely operation at higher beta) and tested the plasma physics in new regimes. However, whilst interesting devices for study, the perceived technological difficulties due to the compact shape initially prevented STs being seriously considered as viable power plants.

Then, in the 2010s, high temperature superconductor (HTS) materials became available as a reliable engineering material, fabricated into long tapes suitable for winding into magnets.  Realising the advantages of this material and its possibilities for fusion, Tokamak Energy proposed a new spherical tokamak path to fusion power and began working on demonstrating the viability of HTS for fusion magnets. The company is now operating a compact tokamak with copper magnets, R₀~0.4m, R/a~1.8, and target I_p=2MA, B_t0=3T, whilst in parallel developing a 5T HTS demonstrator tokamak magnet.  

Here Dr Windridge will discuss why HTS can be a game-changer for tokamak fusion. She will outline Tokamak Energy’s solution for a faster way to fusion and discuss plans and progress, including benefits of smaller devices on the development path and advantages of modularity in power plants. She will indicate some of the key research areas in compact tokamaks and introduce the physics considerations behind the ST approach, to be further developed in the talk by Alan Costley.   

System code and analytical studies have shown that in addition to the conventional large size, high aspect ratio approach to realising fusion power, there could be an approach based on the low aspect ratio spherical tokamak at much smaller size [1]. Small devices would enable accelerated tokamak development because they offer relatively rapid and less expensive development cycles. However, small devices require novel technology and advanced engineering. High temperature superconductors are potentially the enabling technology because they can provide and withstand the necessary high fields used in the toroidal magnets. Other areas are important too, for example plasma start-up and ramp down with a limited, or without, a solenoid, divertor loads, and stresses in the magnet structure. Tokamak Energy is pursuing a development programme that aims to realise this alternative route to fusion power. In this presentation, the physics basis of the approach will be outlined and the key technology and engineering aspects will be highlighted and potential solutions identified. A technology roadmap to deal with the physics and engineering challenges on this path to fusion is under development and will be briefly introduced.

[1] A. E. Costley, J. Hugill and P Buxton, 2015, ‘On the power and size of tokamak fusion pilot plants and reactors’, Nuclear Fusion 55, 033001.