GB2494185A - A spherical tokamak with toroidal field magnets made from high-temperature superconductor - Google Patents

Abstract

An efficient compact nuclear fusion reactor for use as a neutron source or energy source is described. The reactor comprises a toroidal plasma chamber 44 and a plasma confinement system arranged to generate a magnetic field for confining a plasma in the chamber. The plasma confinement system is configured so that a major radius of the confined plasma is 1.5 m or less. The reactor is constructed using High-Temperature Superconducting toroidal magnets 41, which may be operated at low temperature (77K or lower) to provide enhanced performance. The toroidal magnetic field can be increased to 5T or more giving significant increases in fusion output, so that neutron output is very efficient and the reactor can produce a net output of energy.

Description

EFFICIENT COMPACT FUSION REACTOR

Technical Field

The present application relates to a compact fusion reactor operated at high toroidal field. In particular, though not exclusively, the invention relates to a spherical tokamak reactor suitable for use as an energy source or as a highly efficient neutron source Backciround World fusion research has entered a new phase after the beginning of the construction of ITER, the largest and most expensive (clsbn Euros) tokamak ever built. The successful route to a commercial fusion reactor demands long pulse, stable operation combined with the high efficiency required to make electricity production economic.

These three conditions are especially difficult to achieve simultaneously, and the planned programme will require many years of experimental research on ITER and other fusion facilities, as well as theoretical and technological research. It is widely anticipated that a commercial fusion reactor developed through this route will not be built before 2050.

To obtain the fusion reactions required for economic power generation (i.e. much more power out than power in), the conventional tokamak has to be huge (as exemplified by ITER) so that the energy confinement time (which is roughly proportional to plasma volume) can be large enough so that the plasma can be hot enough for thermal fusion to occur.

The challenge of producing fusion power is hugely complex. Many alternative devices apart from tokamaks have been proposed, though none have yet produced any results comparable with the best tokamaks currently operating such as JET.

Sum ma rv In accordance with a first aspect of the present invention there is provided a compact nuclear fusion reactor for use as a neutron (or net energy) source. The reactor comprises a toroidal plasma chamber and a plasma confinement system arranged to generate a magnetic field for confining a plasma in the plasma chamber. The plasma confinement system is configured so that a major radius of the confined plasma is 1.5 m or less, preferably 1.0 m or less, more preferably 0.5 m or less. The plasma confinement system includes HTS coils capable of generating a magnetic field having a toroidal component measured at the major radius of 5 T or more, preferably lOT or more, more preferably 15T or more. The HTS coils may be cooled in use to 77 K, or optionally to 30 K or less or even 4 K or less.

Previous designs for small fusion devices usually have a problem with wall loading -Le. the neutron flux or dispersion of plasma heat through the walls of the plasma chamber. The optional use of a low power input to the plasma of 10 MW or less, preferably 6 MW or less, more preferably 3 MW or less, enables the device to be viable with existing materials and technology. The reactor may be a spherical tokamak.

The neutron output from such a reactor can be at least 1 MW even with conventional copper magnets. When HTS toroidal field magnets are used it will be possible to operate the reactor at considerably higher toroidal field, with large increases in fusion power output.

Neutron production may be enhanced by directing one or more neutral beams into the plasma. The neutral beam or beams may have an energy of less than 200keV, preferably less than 130 key, more preferably less than 80 key, more preferably less than 40 key. Multiple neutral beams may be directed into the beams from directions selected to optimise fusion reactions between particles in the beams.

In one embodiment, the plasma is maintainable in a steady state for more than 10 seconds, preferably more than 100 seconds, more preferably more than 1000 seconds.

This dramatically increases the usefulness of the neutron or energy production, since the total number of neutrons and amount of energy emitted increases with long pulses.

In order to achieve such long pulses, the plasma current may be driven without induction. Lower energy beams can be more efficient (per unit energy input) at transferring momentum to drive the current.

The plasma may be initiated using merging-compression, magnetic pumping so that an oscillating current produces plasma rings to augment the plasma current, activation of one or more solenoids (which may be retractable) located in a central core of the toroidal chamber, and/or Election Bernstein Wave current initiation by a gyrotron. The plasma current may be ramped up using activation of the solenoids, Electron Bernstein Wave current drive, and/or heating the plasma so that a rapid increase in poloidal field necessary to contain the plasma as it grows inputs almost sufficient flux to ramp up the plasma current to a desired working value. If retractable solenoids are used they may include pre-cooled high temperature superconducting solenoids.

The neutral beam(s) and/or plasma may include tritium to enhance neutron production.

Tritium is expensive and radioactive, so it may be preferable to operate the reactor using deuterium only. Some neutrons will still be produced by D-D fusion reaction (it is normally assumed that D-D fusion will produce approximately 1180 as many as produced by D-T fusion under the same conditions of toroidal field, plasma current and plasma heating). However D-D fusion can be important for testing of reactors prior to the use of tritium and in circumstances where the use of tritium is undesirable, eg for reasons of cost, complexity, safety, regulation or availability.

There are certain circumstances where surprisingly high neutron fluxes can be achieved with D-D fusion. This can be achieved by increasing the toroidal field, by judicious use of neutral beam injection and by optimising the methods of plasma heating, andlor by application of ICRH (Ion Cyclotron Resonance Heating) which has been shown to increase neutron production more than tenfold [4].

The neutrons emitted by the reactor may be used, infer a/ia, for generation of electricity, production of heat, formation of isotopes for medical and other use, cancer therapy, production of hydrogen (for example by high temperature electrolysis), treatment of nuclear waste, manufacture of tritium by neutron bombardment of lithium, breeding of nuclear fission fuel, neutron spectroscopy, testing of materials and components, and/or scientific research.

In conventional fusion reactors, a-particles generated in the plasma are retained.

Whereas for the low fields originally proposed for SCFNS the a-particles are not confined, in the higher fields envisaged for the invention described here they will be confined, and will give a significant contribution to the plasma heating.

While the reactor is running, there should optionally be no solenoid in the centre of the torus, since it could be damaged by the high neutron fluence.

The cryostat can be designed with or without liquid cryogens and the cryogens could be a range of molecules or compounds including He, H2, Ne or N2 depending on the temperature and cooling power required. The cryostat can also be designed to add structural strength and rigidity to the tokamak and the toroidal field coils.

The HTS can be manufactured from a range of materials including (Re)BCO in the form of tape or wire with a range of substrates, stabilizers, buffers and overlayers in order to give the structural properties and engineering current density required.

The fusion reactor may comprise divertor plates optimised to reduce the load per unit area on the walls of the plasma chamber, and/or divertor coils configured to direct an exhaust plume of the plasma and expand a footprint of said exhaust plume to large radius and/or sweep the contact region over the exhaust area. One or more of the divertors may be coated with liquid Lithium.

The reactor may also comprise a multiplier blanket configured to increase the flux of emitted neutrons (at the expense of individual neutron energy). Reflector blankets may be provided to direct neutrons out of the reactor in such a way as to produce local increases in flux density and/or to protect poloidal coils and other tokamak components from extensive neutron irradiation.

The reactor may also comprise a sub-critical blanket of fissile or fertile (eg thorium) material forming a hybrid fusion-fission reactor. In this arrangement the copious quantities of neutrons produced by fusion will start and sustain a fission reaction and/or convert fertile isotopes to fissile isotopes. This arrangement can be use to breed new nuclear fuel, destroy nuclear waste and/or generate energy.

Brief Description of the Drawings

Some preferred embodiments of the invention will now be described by way of example only and with reference to the accompanying drawings, in which: Figure 1 shows a comparison of LTS superconductors (left side) with 1st and 2nd generation HTS. The HTS material offers similar performance in a given magnetic field but at higher and more convenient temperatures; Figure 2 shows the critical current as a function of magnetic field for an HTS sample at different temperatures; Figure 3 illustrates the magnetic field line behaviour in conventional and spherical tokamaks; and Figure 4 is a half cross-section through a spherical tokamak.

Detailed Description

Fusion neutrons are produced when a deuterium-tritium (D-T) or deuterium-deuterium (D-D) plasma becomes very hot so that the nuclei fuse together, releasing energetic neutrons. In the conventional (ITER-like) approach to fusion, the plasma needs to have high confinement time, high temperature, and high density to optimise this process.

One way of achieving this is to use a tokamak. A tokamak features a combination of strong toroidal magnetic field BT, high plasma current l and usually a large plasma volume and significant auxiliary heating, to provide a hot stable plasma so that fusion can occur. The auxiliary heating (for example via tens of megawatts of neutral beam injection of high energy H or D or T) is necessary to increase the temperature to sufficiently high values, and/or to maintain the plasma current.

The problem is that, because of the large magnetic fields and high plasma currents generally required, build costs and running costs are high and the engineering has to be robust to cope with the large stored energies present, both in the magnet systems and in the plasma, which has a habit of disrupting' -mega-ampere currents reducing to zero in a few thousandths of a second in a violent instability.

The situation can be improved by contracting the donut-shaped torus of a conventional tokamak to its limit, having the appearance of a cored apple -the spherical' tokamak (ST). The first realisation of this concept at Culham demonstrated a huge increase in efficiency -the magnetic field required to contain a hot plasma can be reduced by a factor of 10. In addition, plasma stability is improved, and build costs reduced.

The major drawback of the ST is that space constraints on the central column prohibit installation of the substantial shielding necessary to protect the central windings in a neutron environment -so conventional toroidal field windings, and conventional central solenoids (used to induce and maintain the plasma currents) are not practical.

However, power plants based on the ST have been designed (using solid copper centre posts with limited shielding, the post being changed every year or so when damaged by neutrons). The drawback with this is that the high energy dissipation in the centre column due to the relatively high resistivity of warm copper, requires a large device for electricity production to become economical.

An important factor is the strength of the toroidal magnetic field, B1.Fusion power from thermal fusion in a tokamak is approximately proportional to the cube of BT and so tokamaks are designed to use the maximum possible Br consistent with the significant stresses this imposes, and the significant costs of electricity required to power these magnets. To minimize these costs long-pulse modern devices such as ITER feature low-temperature superconducting (LTS) magnets cooled by liquid helium. A drawback of the ST approach is that due to the reduced space in the centre column the toroidal field magnet therein is of limited size and so only relatively low toroidal fields have been achieved in STs to date.

The present limit of the high-field approach is exemplified by the medium-sized IGNITOR project, now under development as a joint Russian -Italian project: IGNITOR is predicted to achieve short pulse ignition without need of extensive auxiliary heating, by virtue of its very high field B1, -13 Tesla at the plasma major radius (1.43m) and - 20T at the edge of the centre stack, obtained by conventional copper magnets with a steel support structure.

A smaller scale approach is to use the effect first suggested by Jassby [1] whereby injection of a high energy neutral beam into a small, merely warm', plasma can also produce significant fusion power. This effect combined with an ST, is the basis of our design for a Super Compact Neutron Source' (SCFNS) which has B1-=1.5 Tesla [2].

The power (P18) produced by SCFNS operating with D-T fusion is estimated at 1-2MW, whereas input power (PNBI) is -6MW of NBI; hence 0 (P3 / PNBI) 0.25, although Qeng (PuI Ptotai) is -0.05 since to create 6MW of NBI requires -18MW of electricity; and about a further 10MW is lost in dissipation in the copper magnets ProdUction of net electrical power from fusion requires Qeng>1 Nonetheless SCFNS produces significant fusion power for a small device, and the 14MeV neutrons can have many valuable applications that compensate for the low efficiency of conversion of electrical power input to fusion power output.

Until now it has been thought that this smaller scale approach could not lead to an economic fusion energy power plant, as the input neutral beam injection (NBI) power is relatively large and the magnetic fields are not sufficient to contain the hot, charged alpha particles produced by fusion reactions within the plasma, which therefore loses the self-heating they could provide, and which is a key feature of conventional tokamak designs aimed at fusion power production.

High temperature superconductors Recent advances in high temperature superconductors (HTS) have far-reaching consequences for fusion. Whereas conventional low temperature superconductor (LTS) magnets use temperatures in the liquid helium range (-4K), HTS can give similar results at the more convenient and easier to achieve liquid nitrogen temperature of 77K or even higher as shown in Figure 1.

But the advantages of HTS far exceed cost and convenience. As can be seen from Figure 2, if HTS is actually operated at lower temperatures than 77K, the current-carrying ability is greatly increased (seen by moving vertically in figure 2 at any given applied field), and the conductor can operate in much higher fields (seen by moving horizontally in figure 2). Indeed, Oxford Instruments have recently demonstrated an HTS magnet producing nearly 23T, exceeding the 20T maximum achieved by LTS (actually done by inserting an HTS core into an LTS outer).

The combination of higher maximum field, increased current-carrying capability and reduced complexity of cooling means that very high toroidal field HTS magnets may be possible in the limited space of a Spherical Tokamak core. For example, if 30T is feasible at the edge of the centre column (as suggested from Figure 2), this would give 121 at the major radius of an ST of aspect ratio 1.66 such as SCFNS. Fusion power in a beam-driven device such as SCFNS has been observed to be approximately proportional to B-1-cubed [3] similar to the scaling for thermal fusion. This implies that by increasing BT from 1.5T for the existing SCFNS design to 12T for the high field version described here, the fusion powerwould be increased approximately by 12/1.5 cubed, i.e. by 256; so Q19 -64, Qeng -19; and all in a small device! Fields even higher than 301 may be possible with HIS and would lead to even more efficient neutron and energy generation. An additional benefit is that at this high field, the charged alpha particles produced during the fusion reaction will remain in the plasma, providing significant self-heating and further increasing the efficiency of the reactor.

High Temperature Superconducting technology continues to advance rapidly. The first generation HIS material, BSCCO, was rapidly overtaken by YBCO. As well as the discovery of new HIS materials with fundamentally higher critical fields and critical currents, the engineering performance of existing materials such as YBCO (or, more generally (Re)BCO where Re is a rare earth atom) is rapidly being improved with the result that magnets made from HTS can achieve increasingly high fields from increasingly small conductors. In the present specification, it will be understood that HIS materials include any material which has superconducting properties at temperatures above about 30 K in a low magnetic field.

For an efficient ST fusion neutron or energy source to be practical it is desirable to solve the following problems: * Initiating the plasma current without a conventional central solenoid.

* Ramping up the plasma current to the required value.

* Maintaining the plasma current for a long time with low power input.

* Heating the plasma to produce neutrons at low power input.

* Ensuring that the heat load from the plasma on the divertor regions is tolerable.

* Designing a structure capable of protecting itself against neutron damage, whilst producing a fluence of neutrons for scientific and other applications.

Additionally, use of HIS (especially at low temperatures) for the toroidal field can enable much higher magnetic fields which will make neutron production more efficient, and can provide a net source of energy.

Several options have been considered for an efficient compact fusion reactor operating

at high toroidal field, including:

a. A tokamak with copper or superconducting magnets and an aspect ratio A = 3-4 (A=RIa -ratio of large radius R of the torus to a small radius a); b. A tokamak with superconducting magnets and A = 2; c. A compact spherical tokamak with copper magnets and A=1.5-1.8; d. A compact spherical tokamak with LTS magnets and A=1.5-1.8; e. A compact spherical tokamak with HTS magnets and A=1.5-1.8; Options (a) and (b) are well known, but they can only achieve high efficiency if they are large, comparable in size to ITER. Neutral beam injection is important for current drive in such devices, but does not make much contribution to neutron or energy production.

Option (c) is limited by the current carrying capacity and strength of copper. It is difficult to achieve a high field on a small device with low aspect ratio because of the limited space in the centre of the tokamak. In addition the power and cooling requirements for high currents through copper are very substantial and this effectively prevents long pulse operation.

Option (d) is viable and the performance of LTS (materials such as MgB2, Nb3Sn and NbTi) could be satisfactory although the design of a low temperature cryostat to allow these materials to achieve high field is very challenging.

Option (e) is better than option (d) because the performance of HTS is better that LTS.

HTS can sustain a higher current density at a higher magnetic field and higher temperature than LTS. This means the size of the HTS conductor and surrounding cryostat can be smaller than for LTS. In turn this means that the whole device can be smaller and cheaper while operating at higher magnetic field and hence being more efficient.

The present document focuses on option (e), an Efficient Compact Fusion Reactor (ECFR) based on a compact spherical tokamak with HTS toroidal field coils. This provides the opportunity to operate small spherical tokamak reactors at relatively high toroidal fields of ST and above, potentially providing huge increases in the fusion power produced because fusion power is known to increase with the cube (or more) of the toroidal field, at least in the case of thermal fusion. Whereas the design of SCFNS was calculated to provide a neutron power of 1-2MW for a 6MW power input (and 28MW total power input), if the fusion power increases with the cube of the toroidal field S (measured at the major radius), 30 times more fusion power could be produced at ST.

240 times more at lOT, and 1000 times more at 1ST so that a net energy gain becomes possible even from this compact device. In addition, the energy loss in dissipation in the magnets is greatly reduced if HIS is used.

Before describing the device in detail, it is helpful to consider previous studies of fusion devices based on spherical tokamaks.

Stambaugh et alES] in The Spherical Tokamak Path to Fusion Power' described a family of Spherical Tokamaks (STs) including a Pilot Plant with major radii of R -0.7m plasma current Ip -1OMA, central toroidal field B-1-0 -2.8T) which have significant output (400MW) at an aggressive H-tactor (increase in energy confinement over scaling law for conventional tokamaks) -7 and p (measure of efficiency: the ratio of plasma pressure contained to magnetic field pressure required) -62% and a wall loading of 8MW/m2 (wall assumed to be at radius Ro +2a) and which are designed to produce electricity economically.

Hender et al [6] considered a Component Test Facility (CIF) based on a similarly modest sized ST (S -0.7m, Ip -10.3MA, BTo-3T, fusion output -40MW at a modest H-factor -1.3, DN -2.6 and wall load (at Ro+2a) of -0.75MW/m2) designed to produce sufficient neutron fluence to test Fusion Reactor components.

Wilson et al [7] extended the work of Hender at al to propose a CTF again of A --1.6, designed to consume <1kg of Iritium per year and specifically to aid the fast-track approach to Fusion Power by testing components and materials. Their device has R -0.75m, Ip -8MA, BTo -2.8, H-1.3, PNBI -60MW, and yields Pfus -50MW of which about 25% arises from beam-plasma interactions (discussed further below).

Voss et al [8] developed the Wilson design, increasing the size slightly to R = 0.85m, a = 0.55m, with a slight decrease in current and field to 6.5MA and 2.5T, again assuming H1.3, with PNBI = 44MW and Plus = 35MW.

Dnestrovkij et al [9] provided a DINA code simulation of the Wilson CIF, and find by using a different mix of NBI energies (6MW at 4OkeV and 44MW at l5OkeV) they can provide current ramp -up and, aided by a larger Tritium fraction of 70% (cf 50%) obtain the same fusion output (50MW) but at considerably lower plasma current (5.SMA cf 8MA). Although Tritium is scarce and expensive, the option of using a larger Tritium fraction to obtain the same neutron output but at lower plasma pressure (and hence improved plasma stability) is attractive.

All the above studies employ NBI for current drive (providing heating, in conjunction with a-particle heating -note u-particles have low prompt losses at the high plasma currents employed in the above studies), use well-understood technology (e.g. copper windings), and aspect ratios 1.4 -1.6 (at which sufficient Tritium can be bred without need of a centre-column blanket).

Peng et al [10] proposed a larger CTF with R1.2m, A=1.5, k=3.07, Bt1.1-2.2T, lp=3.4-8.2MA, heating power 15-31MW, bootstrap (self-driven current) fraction &Q5, 3 (ratio of fusion power out to input power) 0.5-2.5, Pfus=7.5-75MW. This CTF also has an option of tritium breeding.

Most recently, Kotchenreuther et al [11] proposed a larger CFNS with 100MW fusion output (Ro = 1.35m, aspect ratio 1.8, BT0 =3.11, Ip = 10-14 MA) using their Super X' divertor to solve the critical divertor thermal load problem. Their device is designed for use either as a CTF, or as the basis of a fusion-fission hybrid, or for development of a pure fusion reactor.

In the present case, the requirements are significantly less demanding than those in the above studies, especially the Stanibaugh et al study which requires long-pulse operation close to stability limits and at high wall-loading to ensure cost-effective electricity production. Hender and Wilson require high neutron flux for long periods to provide sufficient component testing, and operate at high plasma current. In the present proposal, demands can be relaxed. Operating at lower power can still allow a net power output with reduced wall loading and reduced cost of construction.

Iwo more recent studies are particularly relevant.

Galvao et al [12] studied a Multi functional Compact Tokamak Reactor Concept' designed with the same objectives as our present study. They proposed a device of major radius Ro = 1.2 (some 50% larger than MAST and NSTX), with A=1.6, Ip = 5MA, BTo = 3.5T, and obtained a fusion gain (0) -1 for a range of auxiliary heating powers from 5MW to 40MW. Interestingly, at lower powers the maximum Q-1 gain occurs at ever lower densities, whereas bootstrap current increases almost linearly with density-so the higher performance options have the advantage of largest self-driven current. However this study did not consider the additional neutron production provided by beam-plasma interactions.

Kuteev et al [131 specifically addressed the need for a small facility developing up to 10MW of fusion power whilst requiring total auxiliary heating and current drive power c 15MW and total power consumption c 30MW. They re-evaluated the smallest (Ro -0.5m) member of the Stambaugh range but under extremely reduced conditions: Ip- 3MA, BTo -1.35T with a neutron fluence of -3 x 1017 n/s corresponding to a fusion power of -1MW and a neutron load 0.1MW/m2. Modelling shows that neutron production is more than doubled by the beam-on-plasma effect. Importantly for a first pilot device, the build cost was estimated at less than £200M.

Thus rather than operating at high plasma current, it may be possible to employ significant NBI auxiliary heating and enjoy significant neutron production from the NBI beam-on-plasma interactions noted by Jassby [1]. This effect occurs when injected beams slow down in a thermal tokamak plasma, and is effective in the ST plasmas considered here.

The spherical tokamak represents a low aspect ratio version of a conventional tokamak and is a crucial component of the present invention.

The concept of a spherical tokamak (ST) was first introduced by Jassby [14] and later by Peng [15]. At the same time, a small low-aspect ratio tokamak GUTTA was constructed and operated at loffe Institute, Russia, confirming some of unique features of the ST concept. The first demonstration of the main advantages of a spherical tokamak (i.e. high beta, high natural elongation, improved stability and enhanced confinement -H-mode) was on the START device [16] which was operating at Culham Laboratory 1990-1998. START was a small tokamak but achieved normalised plasma pressures 13t -40% (which is still a record for tokamaks). In the ST the aspect ratio A of the plasma column is substantially reduced with respect to conventional tokamak aspect ratio range (M3-4), giving significant improvements in plasma stability. The combination of simple construction, excellent results and high reliability confirmed on more than 15 small and medium sized STs operated during the last decade produce a strong motivation for an ST as the next step in fusion research, and the high performance and small size makes the ST economical both in build cost and in tritium consumption (if D-T operation is desired).

Figure 3 (courtesy of Y-K M Peng) illustrates an effect of aspect ratio reduction. The figure shows the peripheral magnetic field lines in a conventional tokamak 31 and in a spherical tokamak 32. In the conventional tokamak 31, magnetic field lines have comparable lengths in the region of a favourable curvature (inner, high field and stable region) and unfavourable curvature of magnetic field (outer unstable region). In the spherical tokamak 32 the field line path in the inner, stable region is significantly higher than in the outer, unstable region and the field line is generally wrapped onto the central core of the plasma column, where the toroidal magnetic field is highest. As the particle motion in a magnetic trap is bound to the field lines, the most straightforward result of an aspect ratio decrease is an increase in the plasma column magneto-hydrodynamic (MHD) stability. This improved MHD stability permits either a significant increase in the plasma current, or a decrease in the toroidal magnetic field strength; this feature has been exploited in the successful ST experiments, notably START at Culham [16]. The figure shows the plasma column 33 in the START tokamak, with sharp plasma edges, demonstrating the excellent confinement properties (H-mode) obtainable in an ST plasma.

To date, STs have produced good physics performance but so far they have low magnetic fields, low heating power and most of them are short pulse devices. The neutron flux is modest as tritium has not been used, and modelling shows that even if a D-T mix could be employed, neutron yield would be small, mainly because of the low

toroidal field.

The proposed device Efficient Compact Fusion Reactor (ECFR) is the first ST to have high magnetic field, high availability, high neutron fluency, low running costs and the capability to produce net energy over extended periods of time.

Main parameters The ECFR device is a long-pulse spherical tokamak with an elongated plasma, and a double-null divertor. The design objectives are to demonstrate routine steady-state operation in hydrogen (enabling optimisation and any necessary modifications to be made without problems of radioactivity), before proceeding to deuterium-deuterium (DD) and then, if desired, to a deuterium-tritium (DT) mix where considerable neutron fluence would result. The design incorporates optional features (notably shielding/neutron reflectors and a heavy water blanket) which allow control of the neutron output for test purposes.

Standard operation produces significant D-T fusion power for a burn length of longer than 1000 sec which is determined as a "quasi steady-state" for most engineering requirements. The injection of neutral beams of energy up to about 200keV provides the main source of auxiliary power. Electron Bernstein Wave (EBW) heating is also considered.

Start-up and ramp-up It is planned to obtain start-up and ramp-up of the plasma current without use of a large central solenoid because, in the final design, the large neutron fluence may prohibit the use of a conventional central solenoid, as there may be insufficient space for the extensive shielding required to protect the windings.

However a major advantage of the use of a spherical tokamak is that the plasmas (having low aspect ratio and high elongation) have low inductance, and hence large plasma currents are readily obtained -the input of flux from the increasing vertical field necessary to restrain the plasma is also significant at low aspect ratio [17].

Experiments on MAST have demonstrated start-up using a 28GHz 100kW gyrotron (assisted by vertical field ramp) at an efficiency of 0.7A/Watt [18]. A gyrotron fitted to ECFR could have power -1MW and would be predicted to produce a start-up current of-700kA.

An alternative scheme is to use a small solenoid (or pair of upper I lower solenoids) made using mineral insulation with a small shielding (or designed to be retracted before D-T operation begins); it is expected that such a coil would have approximately 25% of the volt-secs output as an equivalent solenoid as used on MAST or NSTX. Initial currents of order 0.5MA are expected. The combination of both schemes would be especially efficient.

A novel development of the retractable solenoid' concept is to use a solenoid wound from HTS, to cool it in a cylinder of liquid Nitrogen outside the tokamak, insert it into the centre tube whilst still superconducting, pass the current to produce the initial plasma, then retract the solenoid before D-T operation. Advantages of using HTS include lower power supply requirements, and the high stresses that can be tolerated by the supported HIS winding.

This initial plasma current will be an adequate target for the lower energy NBI beams, and the heating and current drive they produce will provide current ramp up to the Heating and current drive As previously discussed, it is desirable to obtain a significant fluence of neutrons (say 1MW) at minimum auxiliary heating and minimum current drive, in order to minimise build costs, running costs, and to keep divertor heat loads at tolerable levels.

Recent energy confinement scalings, derived from recent results on both MAST at CCEE and NSTX at Princeton suggest that energy confinement in an ST has a stronger dependence on magnetic field, and a lower dependence on plasma current, than the ITER scalings derived for conventional tokamaks, and hence is improved for

the high field of this design.

Various methods of heating (and current drive) including NBI and a range of radio-frequency (RE) methods may be appropriate. NBI is the most widely used scheme and has the advantages of easy injection into the plasma, and less sensitivity to plasma parameters than most SF methods.

NBI is also the most commonly used method of current drive. Its efficiency depends on many parameters -beam energy, angle of injection, density of plasma. Typically 1MW of NBI may drive 0.1MA of plasma current; and since NBI costs approx £3M per MW, this is a major cost. A potentially helpful feature is the self-driven bootstrap' current, produced in a hot, high energy, plasma, which can account for possibly one-half of the required current. However bootstrap current increases with density, whereas NBI current drive reduces at high density, so a careful optimisation is required.

Thermal load on divertors Some of the energy pumped into a plasma either to heat it or produce current drive emerges along the scrape-off-layer (SOL) at the edge of the plasma, which is directed by divertor coils to localised divertor strike points. The power per unit area here is of critical concern in all fusion devices, and would not normally be acceptable in a small neutron or energy source. However in the present proposal the input power is greatly reduced (of order of a few MW, compared to tens of MW in other designs) so the divertor load is correspondingly reduced. Additional methods are used to reduce the load per unit area further, by a combination of strike-point sweeping; use of the natural divertor' feature observed on START; and use of divertor coils to direct the exhaust plume (as advocated by Peng & Hicks [19]); possibly to expand the footprint to large radius as in the super -X' divertor advocated by Kotschenreuther et al [11]. This latter normally requires large currents in the divertor control coils, as these have to be somewhat removed from the neutron source for protection: however this demand is made tractable here because of the modest plasma current required. Further benefit may be gained by use of a flow of liquid Lithium over the target area which will also be used to pump gases from the vessel, for example in a closed Lithium flow loop.

General outline of this device A cross section of a spherical tokamak suitable for use as an energy or neutron source is shown in Figure 4. The major components of the tokamak are a toroidal field magnet (TF) 41, optional small central solenoid (CS) 42 and poloidal field (PF) coils 43 that magnetically confine, shape and control the plasma inside a toroidal vacuum vessel 44.

The centring force acting on the 0-shaped TE coils 41 is reacted by these coils by wedging in the vault formed by their straight sections. The outer parts of the TF coils 41 and external PF coils are optionally protected from neutron flux by a blanket (which may be 020) and shielding 45. The central part of IF coils, central solenoid and divertor coils are only protected by shielding.

The vacuum vessel 44 may be double-walled, comprising a honey-comb structure with plasma facing tiles, and directly supported via the lower ports and other structures.

Integrated with the vessel are optional neutron reflectors 46 that could provide confinement of fast neutrons which would provide up to 10-fold multiplication of the neutron flux through ports to the outer blanket where neutrons either can be used for irradiation of targets or other fast neutral applications, or thermalised to the low energy to provide a powerful source of slow neutrons. The reason for such assembly is to avoid interaction and capture of slow neutrons in the structures of the tokamak. The outer vessel optionally contains D20 with an option for future replacement by other types of blanket (Pb, salts, etc.) or inclusion of other elements for different tests and studies. The outer shielding will protect the TF and PF coils, and all other outer structures, from the neutron irradiation. The magnet system (TE, PE) is supported by gravity supports, one beneath each TF coil.

Inside the outer vessel, the internal components (and their cooling systems), also absorb radiated heat and neutrons from the plasma and partially protect the outer structures and magnet coils from excessive neutron radiation in addition to D20. The heat deposited in the internal components and in the vessel is ejected to the environment by means of a cooling water system. Special arrangements are employed to bake and consequently clean, in conjunction with the vacuum pumping system, the plasma-facing surfaces inside the vessel by releasing trapped impurities and fuel gas.

The tokamak fuelling system is designed to inject the fuelling gas or solid pellets of hydrogen, deuterium, and tritium, as well as impurities in gaseous or solid form. During plasma start-up, low-density gaseous fuel is introduced into the vacuum vessel chamber by the gas injection system. The plasma progresses from electron-cyclotron-heating and EBW assisted initiation, possibly in conjunction with flux from small retractable solenoid(s), and br a merging-compression' scheme (as used on START and MAST), to an elongated divertor configuration as the plasma current is ramped up.

A major advantage of the ST concept is that the plasmas have low inductance, and hence large plasma currents are readily obtained if required -input of flux from the increasing vertical field necessary to restrain the plasma being significant [18]. Addition of a sequence of plasma rings generated by a simple internal large-radius conductor may also be employed to ramp up the current.

After the current flat top is reached, subsequent plasma fuelling (gas or pellets) together with additional heating leads to a D-T burn with a fusion power in the MW range. With non-inductive current drive from the heating systems, the burn duration is envisaged to be extended above 1000 s and the system is designed for steady-state operations. The integrated plasma control is provided by the PF system, and the pumping, fuelling (H, D, T, and, if required, He and impurities such as N2, Ne and Ar), and heating systems based on feedback from diagnostic sensors.

The pulse can be terminated by reducing the power of the auxiliary heating and current drive systems, followed by current ramp-down and plasma termination. The heating and current drive systems and the cooling systems are designed for long pulse operation, but the pulse duration may be determined by the development of hot spots on the plasma facing components and the rise of impurities in the plasma.

The approach outlined above enables the design of an Efficient Compact Fusion Reactor (ECER) that is much smaller than previous designs of fusion reactors aimed at generating net power, having correspondingly lower construction and operational costs (volume from 1/5 to 1115 of existing designs, magnetic field energy and tritium consumption 10 -100 times lower). The ECFR is an ideal first device to evaluate previously untested areas such as steady-state operation, plasma control, tritium operation, etc whilst producing at least 1MW of fusion neutrons ideal for scientific research, materials tests, production of isotopes for medical and other applications, etc. ECFR is capable of producing net energy over an extended length of time. As such it may be much more than a useful demonstration of fusion technology, it may be the first viable demonstration of a fusion power station.

This design is made possible by a novel combination of new and established techniques over a wide range covering plasma initiation; ramp-up of plasma current; key methods of enhancing neutron production at relatively low current, field and auxiliary heating; use of improved energy confinement; means of varying the neutron energy in a controllable and tunable manner; efficient means of producing steady-state operation; methods of handling the exhaust heat load; special methods of construction, featuring shielding and optional reflectors to both protect coil windings and control the neutron output; and the use of HTS to enable exceptionally high toroidal fields in a small spherical tokamak.

Plasma initiation: methods include merging-compression; magnetic pumping whereby an oscillating current produces plasma rings which augment the plasma current; use of a retractable solenoid, or pair of such solenoids; Electron Bernstein Wave (EBW) current initiation by a gyrotron.

Current ramp-up: methods include retractable solenoid(s), which may be pre-cooled high temperature superconductor solenoids; EBW current drive; and the efficient drive produced by heating the plasma so that the rapid increase in poloidal field necessary to contain the growing plasma inputs almost sufficient flux to ramp up the plasma current to the desired working value.

Enhanced neutron production: in a conventional fusion device nearly all neutron production arises from the central highest temperature region of the plasma. In the proposed device, most neutron production is from interaction of one or more neutral beams with the plasma. In addition, new modelling shows that neutron production is further enhanced by the relatively long path of the NBI beam when directed at optimum angle through the highly-elongated plasma (a natural feature of an SI) and by optimising the Tritium fraction.

Variable neutron energy: in a conventional fusion device the neutron energy is fixed at 14 MeV for D-I fusion and 2.5 MeV for D-D fusion. In one version of the proposed device an antenna configured to induce ion cyclotron resonance heating (ICRH) would be mounted inside the toroidal chamber. This ICRH system could also be configured to increase the energy of the emitted neutrons by several MeV in a controllable and tunable manner.

Optimising neutron output from D-D fusion: while D-T fusion is the best way to achieve the highest neutron flux and energy for some applications, it may be more effective to avoid the problems associated with tritium (eg cost, complexity, safety, regulation or availability) and instead use ICRH to increase neutron energy and/or to heat a D-D plasma to increase neutron flux. This use of ICRH can be combined with higher toroidal field and higher plasma current to give a surprisingly high neutron output from D-D Fusion in a system that may be more cost effective than a D-T Fusion system producing the same neutron flux. Data from JET [4] shows that ICRH can increase the neutron yield by a factor of 14 for D-D fusion.

Favourable confinement scaling: recent research suggests that energy confinement in an ST has a stronger dependence on magnetic field, and a lower dependence on plasma current, than the ITER scalings derived for conventional tokamaks.

Construction features: insulation of the low-voltage Toroidal Field coil segments can be by stainless steel which combines great strength and relatively high resistance; the TF system may be demountable, utilising high-duty versions of the feltmetal sliding joints developed by Voss at CCFE; the device itself could feature a combination of heavy-water tanks and layers of shielding Jreflectors (eg of Be or Pb) to protect the PF coils and external TF coils from lower energy neutrons, and to direct the main stream of neutrons for research and processing tasks.

It is also possible to shoot positive ion beams directly into the plasma through iron

tubes which shield out the magnetic field.

It will be appreciated that compact fusion reactors such as those described herein have a much larger surface area per unit plasma volume than bigger tokamaks. In general costs and implementation difficulty scale at least linearly with plasma volume, while energy output (which can be considered to be limited by acceptable damage levels) scales linearly with surface area. In addition, the costs of a one (or few) of a kind" device are well known to be higher than the costs of many of a kind" devices. It therefore seems likely that many smaller fusion reactors should be cheaper per unit net power output than one large fusion reactor.

It will be appreciated that variations from the above described embodiments may still fall within the scope of the invention.

References [1] .Jassby 0 L Optimisation of Fusion Power Density in the Two Energy Component Tokamak Reactor' Nuclear Fusion 1975 Vol 15 p453 [2] A. Sykes et al, Fusion for Neutrons -a realizable fusion neutron source', Proc of 24 IEEE Symposium on Fusion Enginering, Chicago 2011 Invited Paper 502B-1 [3] M.Valovic et al, Nuclear Fusion 51(2011) 073045 [4] C Hellesen et al, Nuclear Fusion 50 (2010) 022001 [5] R D Stambaugh et al, The Spherical Tokamak Path to Fusion Power', Fusion Technology Vol 33 P1(1998) [6] T C Hender et al, Spherical Tokamak Volume Neutron Source', Fusion Engineering and Design 45 (1990) p265 [1 H R Wilson et al The Physics Basis of a Spherical Tokamak Component Test Facility Proc. 31st EPS Conf 2004 [8] G M Voss et al, Conceptual design of a Component Test Facility based on the Spherical Tokamak'. FED 83 (2008) p1 648 [9] A Dnestrovskij et al, Plasma Devices and Operations, 15, 2007, p1 [10] Y-K M Peng eta/ 2005 Plasma Phys. ControL Fusion 47 B263 [11] M. Kotschenreuther, P. Valanju, S. Mahajan, L.J. Zheng, L.D. Pearlstein, RH.

Bulmer, J. Canik and R. Maingi The super X divertor (SXD) and a compact fusion neutron source (CFNS)'Nucl. Fusion 50(2010)035003 (8pp) [12] R M 0 Galvao et al, Physics and Engineering Basis of a Multi-functional Compact Tokamak Reactor Concept', paper FT/P3-20, IAEA conf 2008 [13] B V Kuteev et al, Plasma and Current Drive parameter options for a low-power Fusion Neutron Source' 23rd IEEE/NPSS Symposium on Fusion Engineering, 2009.

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[14] D L Jassby, Comments Plasma Phys. Controlled Fusion, 3 (1978) 151 [15] Y-K.M. Peng and D.J. Strickler, NucI. Fusion 26, 769 (1986).

[16] M Gryaznevich at all, Phys. Rev. Letters, 80, (1998) 3972 Technology, London, U.K., 3-7 September 1990, Vol 2 p1288 [17] 0. Mitarai and Y. Takase, Plasma current ramp-up by the outer vertical field coils in a spherical tokamak reactor, Fusion Sci. Technol. 43 (2003), [18] V. Shevchenko, Nuclear Fusion Vol 50(2010) p22004 [19] Y-K M Peng and J B Hicks,: proceedings of the 16th Symposium on Fusion

Claims (1)

1. <claim-text>CLAIMS: 1. A compact nuclear fusion reactor comprising a toroidal plasma chamber and a plasma confinement system arranged to generate a magnetic field for confining a plasma in the plasma chamber, wherein: the plasma confinement system is configured so that the major radius of the confined plasma is 1.5 m or less; the plasma confinement system includes toroidal field magnets made from High Temperature Superconductor, preferably cooled in use to 77K, more preferably to 30K or less, more preferably to 4K or less.the magnetic field in use includes a toroidal component of ST or more, preferably lOT or more, more preferably 1ST or more.</claim-text> <claim-text>2. The fusion reactor of claim 1, wherein the major radius of the confined plasma is less than 1.0 m, more preferably less than 0.5 m.</claim-text> <claim-text>3. The fusion reactor of any preceding claim, which reactor is a spherical tokamak reactor.</claim-text> <claim-text>4. The fusion reactor of any preceding claim, arranged to enhance neutron and energy production by directing one or more neutral beams into the plasma.</claim-text> <claim-text>5. The fusion reactor of claim 4, wherein the neutral beams have an energy of less than 200keV, preferably less than 130 key, more preferably less than 80 key, more preferably less than 40 key.</claim-text> <claim-text>6. The fusion reactor of claim 5, wherein the neutral beams include tritium atoms.</claim-text> <claim-text>7. The fusion reactor of claim 5 or 6, wherein the plasma includes tritium ions.</claim-text> <claim-text>8. The fusion reactor of claim 5, wherein the neutral beams include deuterium atoms but not tritium atoms, and the plasma includes deuterium ions but not tritium ions.</claim-text> <claim-text>9. The fusion reactor of any preceding claim, wherein neutral beams are directed into the plasma from different directions selected to optimise fusion reactions between particles in the beams.</claim-text> <claim-text>10. The fusion reactor of any preceding claim, configured so that power input to the plasma is less than 100 MW, preferably less than 10 MW, more preferably less than 6 MW, more preferably less than 3 MW.</claim-text> <claim-text>11. The fusion reactor of any preceding claim, arranged to operate at a fusion energy gain factor Qeflg>l, more preferably Qeflg>3, more preferably Qeflg>lO, more preferably Qeng>lS, more preferably Qeng>20, and operated either as an efficient neutron source or an energy source.</claim-text> <claim-text>12. The fusion reactor of any preceding claim, wherein the plasma is maintainable in a steady state for more than 10 seconds, preferably more than 100 seconds, more preferably more than 1000 seconds.</claim-text> <claim-text>13. The fusion reactor of claim 12, wherein the plasma current is driven without induction.</claim-text> <claim-text>14. The fusion reactor of claim 13, arranged to initiate the plasma using one or more of the following operations: merging-compression: magnetic pumping so that an oscillating current produces plasma rings to augment the plasma current; activation of one or more solenoids (which may be retractable) located in a central core of the toroidal chamber; and Electron Bernstein Wave current initiation by a gyrotron.</claim-text> <claim-text>15. The fusion reactor of claim 14, arranged to ramp up the plasma current using one or more of the following operations: activation of the one or more solenoids (which may be retractable): Electron Bernstein Wave current drive; and heating the plasma so that a rapid increase in poloidal field necessary to contain the plasma as it grows inputs almost sufficient flux to ramp up the plasma current to a desired working value.</claim-text> <claim-text>16. The fusion reactor of claim 14 or 15, wherein the one or more retractable solenoids include one or more pre-cooled high temperature superconducting solenoids.</claim-text> <claim-text>17. The fusion reactor of any preceding claim, arranged to supply an output of neutrons of at least 1 MW, more preferably at least 10MW, more preferably at least 100MW.</claim-text> <claim-text>18. The fusion reactor of any preceding claim in which the cryostat casing and/or the HTS manufactured material is configured to provide at least some structural integrity 19 The fusion reactor of any preceding claim in having Poloidal Field Coils made of High Temperature Superconductor 20. The fusion reactor of any preceding claim, wherein the output neutrons are usable for one or more of: production of electrical energy formation of isotopes for medical and other use; production of hydrogen: production of heat; treatment of nuclear waste; manufacture of tritium by neutron bombardment of lithium; breeding of nuclear fission fuel; driving a sub-critical fission or fertile blanket in a fusion-fission hybrid materials analysis including neutron spectroscopy and/or neutron imaging (tomography) and/or neutron activation analysis; materials processing by neutron irradiation detection of clandestine materials medical imaging medical therapy including neutron capture therapy and/or neutron beam therapy testing of materials and components; and scientific research.21. The fusion reactor of any preceding claim, wherein the plasma confinement system is configured so that a-particles generated in the plasma are confined.22. The fusion reactor of any preceding claim, wherein the plasma confinement system is configured so that no solenoid is located in the centre of the toroidal plasma chamber when the reactor is in operation to fuse deuterium and tritium.23. The fusion reactor of any preceding claim, further comprising divertors optimised to reduce the load per unit area on the walls of the plasma chamber.24. The fusion reactor of claim 23, further comprising divertor coils configured to direct an exhaust plume of the plasma and expand a footprint of said exhaust plume to large radius and/or major radius and/or sweep the contact region over the divertors.25. The fusion reactor of claim 23 or 24, wherein part or all of the surface of the divertors is coated with Lithium.26. The fusion reactor of any preceding claim, further comprising an antenna configured to induce ion cyclotron resonance heating (ICRH) and configured to increase the energy of the emitted neutrons in a controllable and tunable manner.27. The fusion reactor of any preceding claim, wherein the vertical elongation of the confined plasma at a separatrix that separates a core plasma and a region of openmagnetic field lines is about 3.28. The fusion reactor of any preceding claim, further comprising a multiplier blanket configured to increase the flux of emitted neutrons.29. The fusion reactor of any preceding claim, further comprising reflectors to direct neutrons out of the reactor so as to produce a local increase in flux density.30. A method of generating neutrons or energy by operating a nuclear fusion reactor comprising a toroidal plasma chamber, the method comprising: initiating a plasma in the plasma chamber; generating a magnetic field with a toroidal component of of ST or more, preferably 10 T or more, more preferably 15 1 or more to confine the plasma in the plasma chamber, the plasma having a major radius of 1.5 m or less; emitting neutrons.31. The method of claim 30, wherein the major radius of the confined plasma is less than 1.0 m, more preferably less than 0.5 m.32. The method of claim 30 or 31, further comprising inputting energy to the plasma at less than 100 MW, preferably less than 10 MW, more preferably less than 6 MW, more preferably less than 3 MW.33. The method of any of claims 30 to 32, wherein the plasma operates at a fusion energy gain factor Qeng>i, more preferably Qeng>3, more preferably Qeng>10 more preferably Qeng>15' more preferably °eng>20- 34. The method of any of claims 30 to 33, further comprising maintaining the plasma in a steady state for at least 10 seconds, preferably at least 100 seconds, more preferably at least 1000 seconds.35. The method of any of claims 30 to 34, wherein the plasma is initiated using one or more of the following operations: merging-compression; magnetic pumping so that an oscillating current produces plasma rings to augment the plasma current; activation of a one or more solenoids (which may be retractable) located in a central core of the toroidal chamber; and Electron Bernstein Wave current initiation by a gyrotron.36. The method of any of claims 30 to 35, wherein the plasma current is ramped up using one or more of the following operations: activation of the one or more solenoids (which may be retractable); Electron Bernstein Wave current drive; and heating the plasma so that a rapid increase in poloidal field necessary to contain the plasma as it grows inputs almost sufficient flux to ramp up the plasma current to a desired working value.37. The method of claim 35 or 36, wherein the one or more retractable solenoids include one or more pro-cooled high temperature superconducting solenoids.38. The method of any of claims 30 to 37, further comprising directing one or more neutral beams having an energy of less than 200keV, preferably less than 130 key, more preferably less than 80 key, more preferably less than 40 key beam, into the plasma so as to enhance neutron production.39. The method of any of claims 30 to 38, wherein the neutrons are generated at a rate of at least 3 x 1017 neutrons per second, preferably at least lola neutrons per second, more preferably at least 1019 neutrons per second, more preferably at least 1020 neutrons per second.40. The method of any of claims 30 to 39, wherein at least one of the neutral beams and the plasma comprises tritium.41. The method of any of claims 30 to 39, wherein the neutral beams and plasma each comprise deuterium only so that fusion operates as a D-D reaction.42. The method of any of claims 30 to 41, further comprising using the neutrons for one or more of: production of electrical energy formation of isotopes for medical and other use; production of hydrogen; production of heat treatment of nuclear waste; manufacture of tritium by neutron bombardment of lithium; breeding of nuclear fission fuel; driving a sub-critical fission or fertile blanket in a fusion-fission hybrid materials analysis including neutron spectroscopy and/or neutron imaging (tomography) and/or neutron activation analysis; materials processing by neutron irradiation detection of clandestine materials medical imaging medical therapy including neutron capture therapy and/or neutron beam therapy testing of materials and components; and scientific research.</claim-text>