(19) TZZ Z_T

(11) EP 2 752 099 B1

(12) EUROPEAN PATENT SPECIFICATION

(45) Date of publication and mention (51) Int Cl.: of the grant of the patent: H05H 1/12 (2006.01) G21B 1/05 (2006.01) 05.07.2017 Bulletin 2017/27 (86) International application number: (21) Application number: 12759807.6 PCT/GB2012/052093

(22) Date of filing: 24.08.2012 (87) International publication number: WO 2013/030554 (07.03.2013 Gazette 2013/10)

(54) EFFICIENT COMPACT FUSION REACTOR KOMPAKTER EFFIZIENTER FUSIONSREAKTOR RÉACTEUR DE FUSION COMPACT EFFICACE

(84) Designated Contracting States: • Galvão et al.: "Physics and Engineering Basis of AL AT BE BG CH CY CZ DE DK EE ES FI FR GB Multi-functional Compact Tokamak Reactor GR HR HU IE IS IT LI LT LU LV MC MK MT NL NO Concept", Proceedings of the 22nd IAEA Fusion PL PT RO RS SE SI SK SM TR Energy Conference - Geneva, 13-18 October 2008 , FT/P3-20, 27 March 2009 (2009-03-27), (30) Priority: 02.09.2011 GB 201115188 pages 1-8, XP002691321, Retrieved from the 10.07.2012 GB 201212241 Internet: URL:http://www-naweb.iaea.org/napc/physics (43) Date of publication of application: /FEC/FEC2008/papers/ft_p3-20.pdf [retrieved on 09.07.2014 Bulletin 2014/28 2013-01-28] • SYKES A ED - SYKES A: "The Development of (73) Proprietor: Tokamak Energy Ltd the SphericalTokamak", INTERNET CITATION, 1 Abingdon September 2008 (2008-09-01), pages 1-62, Oxfordshire OX14 3DB (GB) XP002655495, Retrieved from the Internet: URL:http://www.triam.kyushu-u.ac.jp/ICPP/p (72) Inventors: rogram/download/12-PL01.pdf [retrieved on • KINGHAM, David 2011-08-02] Oxford • Reiersen et al.: "The engineering design of TPX", Oxfordshire OX2 9JG (GB) 15th IEEE/NPSS Symposium. Fusion •SYKES,Alan Engineering, 11-15 Oct. 1993, Hyannis, MA, USA Wantage 15th IEEE/NPSS Symposium. Fusion Oxfordshire OX12 9AS (GB) Engineering (Cat. No.93CH3348-0), vol. 1 1994, • GRYAZNEVICH, Mikhail pages 387-392, XP002691322, USA DOI: Abingdon 10.1109/FUSION.1993.518356 ISBN: Oxfordshire OX14 4NA (GB) 0-7803-1412-3 Retrieved from the Internet: URL:http://ieeexplore.ieee.org/stamp/stamp .jsp (74) Representative: Talbot-Ponsonby, Daniel ?tp=&arnumber=518356 [retrieved on Frederick 2013-01-30] Marks & Clerk LLP Fletcher House Heatley Road The Oxford Science Park Oxford OX4 4GE (GB)

(56) References cited: WO-A1-2011/154717 US-A- 4 370 296

Note: Within nine months of the publication of the mention of the grant of the European patent in the European Patent Bulletin, any person may give notice to the European Patent Office of opposition to that patent, in accordance with the Implementing Regulations. Notice of opposition shall not be deemed to have been filed until the opposition fee has been paid. (Art. 99(1) European Patent Convention). EP 2 752 099 B1

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• NAGAYAMA Y ET AL: "Proposal of steady state • BUSH C E ET AL: "Combined H-modes in DD and superconducting spherical Tokamak DT plasmas in TFTR", PLASMA PHYSICS AND experiment", DENKI GAKKAI RONBUNSHI. A, CONTROLLED FUSION, IOP, BRISTOL, GB, vol. KISO, ZAIRYO, KYOTSU BUMONSHI 38, no. 8, 3 August 2011 (2011-08-03) , pages -TRANSACTIONS OF THE INSTITUTE OF 1353-1357, XP002655504, ISSN: 0741-3335, DOI: ELECTRICAL ENGINEERS OF JAPAN. A, DENKI 10.1088/0741-3335/38/8/036 GAKKAI, TOKYO, JP, vol. 125, no. 11, 1 January 2005 (2005-01-01), pages 964-965, XP008159727, ISSN: 0385-4205, DOI: 10.1541/IEEJFMS.125.964

2 1 EP 2 752 099 B1 2

Description [0006] The high temperature superconductor (HTS) coils may be capable of generating a toroidal magnetic Technical Field field measured at the major radius of 10T or more. [0007] Previous designs for small fusion devices usu- [0001] The present application relates to a compact 5 ally have a problem with wall loading -i.e. the neutron fusion reactor operated at high toroidal field. In particular, flux or dispersion of plasma heat through the walls of the though not exclusively, the invention relates to a spher- plasma chamber. The optional use of a low power input ical tokamak reactor suitable for use as an energy source to the plasma of 10 MW or less, preferably 6 MW or less, or as a highly efficient neutron source more preferably 3 MW, more preferably 1 MW or less, 10 enables the device to be viable with existing materials Background and technology. [0008] The power output from such a reactor can be [0002] The challenge of producing fusion power is at least 1 MW even with conventional copper magnets. hugely complex. Many alternative devices apart from When HTS toroidal field magnets are used it will be pos- tokamaks have been proposed, though none have yet 15 sibleto operatethe reactor at considerably highertoroidal produced any results comparable with the best tokamaks field, with large increases in fusion power output. currently operating such as JET. [0009] Neutron production may be enhanced by direct- [0003] World fusion research has entered a new phase ing one or more neutral beams into the plasma. The neu- afterthe beginning of theconstruction ofITER, the largest tral beam or beams may have an energy of less than and most expensive (c15bn Euros) tokamak ever built. 20 200keV, preferably less than 130 keV, more preferably The successful route to a commercial fusion reactor de- less than 80 keV, more preferably less than 40 keV. Mul- mands long pulse, stable operation combined with the tiple neutral beams may be directed into the plasma from high efficiency required to make electricity production directions selected to optimise fusion reactions between economic. These three conditions are especially difficult particles in the beams and the thermal plasma, and may to achieve simultaneously, and the planned programme 25 include colliding beams. will require manyyears of experimental research on ITER [0010] In one embodiment, the plasma is maintainable and other fusion facilities, as well as theoretical and tech- in a steady state for more than 10 seconds, preferably nological research. It is widely anticipated that a com- more than 100 seconds, more preferably more than 1000 mercial fusion reactor developed through this route will seconds, more preferably more than 10000 seconds. In- not be built before 2050. 30 deed, the plasma may be maintainable in a steady state [0004] To obtain the fusion reactions required for eco- continuously up to a few years. This dramatically increas- nomic power generation (i.e. much more power out than es the usefulness of the neutron or energy production, power in), the conventional tokamak has to be huge (as since the total number of neutrons and amount of energy exemplified by ITER) so that the energy confinement time emitted increases with long pulses. In order to achieve (which is roughly proportional to plasma volume) can be 35 such long pulses, the plasma current may be driven with- large enough so that the plasma can be hot enough for out induction for example by using neutral beams or RF thermal fusion to occur. Galvão et al.: "Physics and En- current drive. RF current drive includes any electromag- gineering Basis of Multi-functional Compact Tokamak netic wave technique to drive the current including Elec- Reactor Concept", Proceedings of the 22nd IAEA Fusion tron Bernstein Wave, Lower Hybrid and Electron Cyclo- Energy Conference - Geneva, 13-18 October 2008,40 tron Resonance and any combination thereof. Lower en- FT/P3-20, 27, March 2009, describes a smaller tokamak. ergy neutral beams can be more efficient (per unit energy input) at transferring momentum to drive the current. The Summary use of HTS magnets helps to maintain the plasma in a steady state because, being a superconductor, there is [0005] In accordance with a first aspect of the present 45 no heating effect from resistance in the magnet and the invention there is provided a compact spherical tokamak current supplies for HTS magnets are more stable than nuclear fusion reactor for use as an energy source or a power supplies for resistive magnets. highly efficient neutron source. The reactor comprises a [0011] The plasma may be initiated using merging- toroidal plasma chamber and a plasma confinement sys- compression, or magnetic pumping whereby an oscillat- tem arranged to generate a magnetic field for confining 50 ing current produces plasma rings to augment the plasma a plasma in the plasma chamber. The plasma confine- current, or activation of one or more solenoids (which ment system includes toroidal field magnets made from may be retractable) located in a central core of the toroi- material comprising high temperature superconductor dal chamber, and/or RF current initiation by a gyrotron cooled in use to 30K or less and configured such that the or other suitable high frequency generator. The plasma magnetic field in use includes a toroidal component of 55 current may be ramped up using activation of the sole- 5T or more, thereby allowing the plasma confinement noids, RF current drive, and/or heating the plasma so system to be configured so that a major radius of the that a rapid increase in poloidal field necessary to contain confined plasma is 1.5 m or less. the plasma as it grows inputs almost sufficient flux to

3 3 EP 2 752 099 B1 4 ramp-up the plasma current to a desired working value. of heat, formation of isotopes for medical and other use, If retractable solenoids are used they may optionally be cancer therapy, production of hydrogen (for example by pre-cooled high temperature superconducting solenoids. high temperature electrolysis), treatment of nuclear The plasma current may be maintained using RF current waste, manufacture of tritium by neutron bombardment drive and/or Neutral Beam injection. 5 of lithium, breeding of nuclear fission fuel, neutron spec- [0012] The neutral beam(s) and/or plasma may include troscopy, testing of materials and components, and/or tritium to enhance neutron production. Tritium is expen- scientific research. sive and radioactive, so it may be preferable to operate [0019] In conventional fusion reactors, α-particles gen- the reactor using deuterium only. Some neutrons will still erated in the plasma are retained. Although the invention be produced by D-D fusion reaction (it is normally as- 10 described here is much smaller than a conventional fu- sumed that D-D fusion will produce approximately 1/80 sion reactor, the α-particles will still be confined because as many as produced by D-T fusion under the same con- of the high field, and will give a significant contribution to ditions of toroidal field, plasma current and plasma heat- the plasma heating. ing). However D-D fusion can be important for testing of [0020] While the reactor is running, there should op- reactors prior to the use of tritium and in circumstances 15 tionally be no solenoid in the centre of the torus, since it where the use of tritium is undesirable, eg for reasons of could be damaged by the high neutron fluence. cost, complexity, safety, regulation or availability. [0021] A feature of the present invention is that High [0013] There are certain circumstances where surpris- Temperature Superconductor (HTS) is used in the main inglyhigh neutron fluxes canbe achieved with D-D fusion. toroidal field magnet (and optionally in the other mag- This can be achieved by increasing the toroidal field, by 20 nets), enabling high fields to be obtained at low opera- judicious use of neutral beam injection and by optimising tional cost in a compact ST. The combination of high field, the methods of plasma heating, and/or by application of small size and low aspect ratio (which provides improved ICRH (Ion Cyclotron Resonance Heating) which has stability and improved energy confinement) enables fu- been shown to increase neutron production more than sion energy to be realisable on a much smaller scale than tenfold [4]. 25 in previous designs. [0014] Shielding may be provided around the central [0022] The HTS cryostat can be designed with or with- column in order to reduce or eliminate damage from neu- out liquid cryogens and the cryogens could be a range trons. The HTS manufactured material may be config- of molecules or compounds including He, H2, Ne or N2 ured to provide enhanced resistance to neutron damage, depending on the temperature and cooling power re- for example by increasing the thickness of the HTS layer 30 quired. The cryostat can also be designed to add struc- within the HTS manufactured material. tural strength and rigidity to the tokamak and the toroidal [0015] The HTS manufactured material may be con- field coils. figured to provide an increased current density, for ex- [0023] The HTS can be manufactured from a range of ample by reducing the thickness of the non-HTS layers materials including YBCO or (Re)BCO (where Re is a or increasing the thickness of the HTS layers within the 35 rare earth element) in the form of tape or wire with a range HTS manufactured material, in order to allow more space of substrates, stabilizers, buffers and overlayers in order for shielding. to give the structural properties and engineering current [0016] The central column may comprise beryllium, density required. aluminium or another non-HTS material that will maintain [0024] The fusion reactor may include divertor plates acceptable levels of structural integrity and conductivity 40 optimised to reduce the load per unit area on the walls despite the neutron flux. The beryllium, aluminium or oth- of the plasma chamber, and/or divertor coils configured er non-HTS material is optionally cryogenically cooled to to direct an exhaust plume of the plasma and expand a reduce its resistance and is optionally joined to HTS ma- footprint of said exhaust plume to large radius and/or terial forming the remainder of the toroidal magnet apart sweep the contact region over the exhaust area. One or from the central column. 45 more of the divertors may be coated with liquid lithium. [0017] The inner part of the central column may be The walls of the vacuum chamber may also be coated made of HTS and the outer part made of beryllium, alu- with liquid lithium. miniumor anothernon-HTS material thatprovides shield- [0025] The reactor may also comprise a multiplier blan- ing against damage to the HTS from neutrons. The be- ket configured to increase the flux of emitted neutrons ryllium,aluminium or other non-HTS material isoptionally 50 (at the expense of individual neutron energy). Reflector cryogenically cooled to reduce its resistance and is op- blankets may be provided to direct neutrons out of the tionally joined to HTS material forming the remainder of reactor in such a way as to produce local increases in the toroidal magnet apart from the central column. The flux density and/or to protect poloidal coils and other toka- HTS material may be configured to provide enhanced mak components from extensive neutron irradiation. resistance to neutron damage and/or enhanced current 55 [0026] The reactor may also comprise a sub-critical density. blanket of fissile or fertile (eg thorium) material forming [0018] The neutrons emitted by the reactor may be a hybrid fusion-fission reactor. In this arrangement the used, inter alia, for generation of electricity, production copious quantities of neutrons produced by fusion will

4 5 EP 2 752 099 B1 6 start and sustain a fission reaction and/or convert fertile um-tritium (D-T) or deuterium-deuterium (D-D) plasma isotopes to fissile isotopes. This arrangement can be use becomes very hot so that the nuclei fuse together, re- to breed new nuclear fuel, destroy nuclear waste and/or leasing energetic neutrons. To date, the most promising generate energy. way of achieving this is to use a tokamak; in the conven- [0027] The invention also provides a power station 5 tional tokamak approach to fusion (as embodied by IT- comprising a plurality of fusion reactors as described ER), the plasma needs to have high confinement time, above. high temperature, and high density to optimise this proc- [0028] In accordance with another aspect of the ess. present invention there is provided a method of generat- [0032] A tokamak features a combination of strong 10 ing neutrons or energy by operating a spherical tokamak toroidal magnetic field BT, high plasma current p I and nuclear fusion reactor comprising a toroidal plasma usually a large plasma volume and significant auxiliary chamber. The method comprises initiating a plasma in heating, to provide a hot stable plasma so that fusion can the plasma chamber, using toroidal field coils comprising occur. The auxiliary heating (for example via tens of meg- high temperature superconductor material cooled to 30K awatts of neutral beam injection of high energy H, D or or less to generate a magnetic field with a toroidal com- 15 T) is necessary to increase the temperature to the suffi- ponent of 5T or more, preferably 10 T or more, more ciently high values required for nuclear fusion to occur, preferably 15 T or more to confine the plasma in the plas- and/or to maintain the plasma current. ma chamber, the plasma having a major radius of 1.5 m [0033] The problem is that, because of the large size, or less, and emitting neutrons and other energetic parti- large magnetic fields, and high plasma currents generally cles. 20 required, build costs and running costs are high and the engineering has to be robust to cope with the large stored Brief Description of the Drawings energies present, both in the magnet systems and in the plasma, which has a habit of ’disrupting’ - mega-ampere [0029] Some preferred embodiments of the invention currents reducing to zero in a few thousandths of a sec- will now be described by way of example only and with 25 ond in a violent instability. reference to the accompanying drawings, in which: [0034] The situation can be improved by contracting the donut-shaped torus of a conventional tokamak to its Figure 1 shows a comparison of Low Temperature limit, having the appearance of a cored apple - the ’spher- Superconductor (LTS) (left side) with 1st and 2nd ical’ tokamak (ST). The first realisation of this concept at generation HTS. The HTS material offers similar per- 30 Culham demonstrated a huge increase in efficiency - the formance in a given magnetic field but at higher and magnetic field required to contain a hot plasma can be more convenient temperatures; however if cooled to reduced by a factor of 10. In addition, plasma stability is low temperatures (e.g. the 4K used for LTS) it can improved, and build costs reduced. carry substantially higher current than LTS. [0035] A drawback of the ST is that the limited space Figure 2 shows the critical current as a function of 35 in the central column prohibits installation of the substan- magnetic field for an HTS sample at different tem- tial shielding necessary to protect the central windings in peratures; a neutron environment - so conventional toroidal field Figure 3 illustrates the magnetic field line behaviour windings, and conventional central solenoids (used to in conventional and spherical tokamaks; induce and maintain the plasma currents) are not prac- Figure 4 is a half cross-section through a spherical 40 tical. Although power plants based on the ST have been tokamak with conventional copper magnets; designed (using solid copper centre posts with limited Figure 5 shows the structure of one example of HTS shielding, the post being changed every year or so when material; and damaged by neutrons), these have high energy dissipa- Figure 6 shows a quarter cross section through a tion in the centre column due to the relatively high resis- spherical tokamak with HTS toroidal field magnets 45 tivity of warm copper, requiring a large device for elec- with limited neutron shielding and different configu- tricity production to become economical. rations of the central column to provide more resil- [0036] An important factor is the strength of the toroidal

ience to neutron bombardment magnetic field, BT. Fusion power from thermal fusion in a tokamak is proportional to the fourth power of BT and Detailed Description 50 so tokamaks are designed to use the maximum possible BT consistent with the significant stresses this imposes, [0030] The present application is based on a very com- and the significant costs of electricity required to power pact form of the tokamak, and employs a range of inno- these magnets. To minimize these costs, long-pulse vative features, including use of High Temperature Su- modern devices such as ITER feature LTS magnets perconducting magnets. The ’Efficient Compact Fusion 55 cooled by liquid helium. Reactor’ (ECFR) is intended to provide a compact fusion [0037] The present limit of the high-field approach is power plant. exemplified by the medium-sized IGNITOR project, now [0031] Fusion neutrons are produced when a deuteri- under development as a joint Russian - Italian project:

5 7 EP 2 752 099 B1 8

IGNITOR is predicted to achieve short pulse ignition with- the conductor can operate in much higher fields (seen out need of extensive auxiliary heating, by virtue of its by moving horizontally in figure 2). Indeed, Oxford Instru- very high field B T, ∼ 13 Tesla at the plasma major radius ments have recently demonstrated an HTS magnet pro- (1.43m) and ∼ 20T at the edge of the centre stack, ob- ducing nearly 23T, exceeding the 20T maximum tained by conventional copper magnets with a steel sup- 5 achieved by LTS (actually done by inserting an HTS core port structure. into an LTS outer). [0038] A drawback of the ST approach is that due to [0044] The combination of higher maximum field, in- the reduced space in the centre column the toroidal field creased current-carrying capability and reduced com- magnet therein is of limited size and so only relatively plexity of cooling means that very high toroidal field HTS low toroidal fields of less than 1 Tesla have been10 magnets may be possible in the limited space of a Spher- achieved in STs to date. This problem is overcome in ical Tokamak core. For example, if 30T is feasible at the ECFR by use of High Temperature Superconducting edge of the centre column (as suggested from Figure 2), magnets. this would give 12T at the major radius of an ST of aspect [0039] A smaller scale approach to fusion is to use the ratio 1.66 such as SCFNS. Fusion power in a beam-driv- effect first suggested by Jassby [1] whereby injection of 15 en device such as SCFNS has been observed to be ap- a high energy neutral beam into a small, merely ’warm’, proximately proportional to B T cubed [3]. This implies that plasma can also produce significant fusion power. This by increasing B T from 1.5T for the existing SCFNS design effect combined with an ST, is the basis of our design for to 12T for the high field version described here, the fusion a ’Super Compact Neutron Source’ (SCFNS) which has power would be increased approximately by 12/1.5 20 BT=1.5 Tesla [2]. cubed, i.e. by 512; so Qfus ∼ 128, Qeng ∼ 38; and all in a [0040] The power (P fus)produced by SCFNS operating small device! An additional benefit is that at these high with D-T fusion is estimated at 1-2MW, whereas input fields, the charged alpha particles produced during the power (PNBI) is ∼6MW of NBI; hence Q (Pfus / PNBI) ∼ fusion reaction will remain in the plasma, providing sig- 0.25, although Q eng (Pfus / P total) is ∼ 0.05 since to create nificant self-heating and further increasing the efficiency 6MW of NBI requires -18MW of electricity; and about a 25 of the reactor. Work by Jassby [1] showed that there is further 10MW is lost in dissipation in the copper magnets. a fundamental limit to the efficiency of idealised beam- Production of net electrical power from fusion requires plasma fusion at around Qfus ∼ 3 so while this would still Qeng>1. Nonetheless SCFNS produces significant fusion allow a highly efficient neutron source from a small de- power for a small device, and the 14MeV neutrons can vice, it is not a viable approach for energy production. have many valuable applications that compensate for the 30 However at the high fields now envisaged we obtain high- low efficiency of conversion of electrical power input to er confinement, higher plasma temperatures and hence fusion power output. a combination of beam plasma fusion and thermal fusion, [0041] Until now it has been thought that this smaller possibly purely thermal fusion without need for neutral scale approach could not lead to an economic fusion en- beam heating. ergy power plant, as the input neutral beam injection35 [0045] The maximum achievable thermal fusion power (NBI) power is relatively large and the magnetic fields is well known to be proportional to the 4 th powerof toroidal 2 4 are not sufficient to contain the hot, charged alpha par- field [5]. In fact it is proportional to β BT V where β is the ticles produced by fusion reactions within the plasma, normalized plasma pressure and V is the volume. The β which therefore loses the self-heating they could provide, limit is 4 to 5 times higher in a spherical tokamak than in 40 and which is a key feature of conventional tokamak de- a conventional low aspect ratio tokamak, so if B T can be signs aimed at fusion power production. However recent as high as 12T or more, and high plasma pressures can advances in technology may enable these small STs to be obtained, then high thermal fusion power is possible achieve high magnetic field, as described below. even from a small spherical tokamak. For example, ITER is expected to produce 500MW of fusion power with a High temperature superconductors 45 toroidal field of 5.5T. A spherical tokamak with twice the toroidal field and 4 times higher β should be able to pro- [0042] Recent advances in high temperature super- duce the same power in 1/256 of the volume. conductors (HTS) have far-reaching consequences for [0046] High Temperature Superconducting technolo- fusion. Whereas conventional low temperature super- gy continues to advance rapidly. The first generation HTS conductor (LTS) magnets use temperatures in the liquid 50 material, BSCCO, was rapidly overtaken by YBCO. As helium range (∼ 4K), HTS can give similar results at the well as the discovery of new HTS materials with funda- more convenient and easier to achieve liquid nitrogen mentally higher critical fields and critical currents, the en- temperature of 77K or even higher as shown in Figure 1. gineering performance of existing materials such as YB- [0043] But the advantages of HTS far exceed cost and CO (or, more generally (Re)BCO where Re is a rare earth convenience. As can be seen from Figure 2, if HTS is 55 atom) is rapidly being improved with the result that mag- actually operated at lower temperatures than 77K, the nets made from HTS can achieve increasingly high fields current-carrying ability is greatly increased (seen by mov- from increasingly small conductors. In the present spec- ing vertically in figure 2 at any given applied field), and ification, it will be understood that HTS materials include

6 9 EP 2 752 099 B1 10 any material which has superconducting properties at the first, second and fourth means, could be used. temperatures above about 30 K in a low magnetic field. [0053] For an efficient ST fusion neutron or energy [0047] The performance of HTS under intense high en- source to be practical it is desirable to solve the following ergy neutron bombardment is not yet known, however problems: there are concerns that it will need more than 10cm of 5 shielding in order to remain effective for months or years • Initiating the plasma current without a conventional of operation. This amount of shielding may be too large central solenoid. to accommodate around the central column of a small • Ramping up theplasma current to therequired value. spherical tokamak. Several alternative means may be • Maintaining the plasma current for a long time with utilized to allow a high current to pass through the central 10 low power input. column. • Heating the plasma to produce neutrons at low pow- [0048] Figure 5 is a schematic illustration of the com- er input. ponents of standard HTS tape 500. Such tape 500 is • Ensuring that the heat load from the plasma on the generally approximately 100 microns thick, and includes divertor regions is tolerable. an electropolished hasteloy substrate 501 approximately 15 • Designing a structure capable of protecting itself 50 microns thick, on which is deposited by IBAD or mag- against neutron damage, whilst producing a fluence netron sputtering a series of buffer stack layers 502, each of neutrons for energy production or for scientific and approximately 0.2 microns thick. An epitaxial (RE)BCO- other applications. HTSlayer 503 (depositedby MOCVD) overlays the buffer layer, and is typically 1 micron thick. A two micron silver 20 [0054] Several options have been considered for an layer 504 is deposited on the HTS layer by sputtering, efficient compact fusion reactor operating at high toroidal and 20 micron copper stabilizer layers 505 are electro- field, including: plated onto both sides of the tape. In order to increase the current in the tape, the thickness of the HTS layer a. A tokamak with copper or superconducting mag- may be increased from around 1 micron to between 4 25 nets and an aspect ratio A = 3-4 (A=R/a - ratio of and 20 microns. This increases the current that can be large radius R of the torus to a small radius a); carried by a factor of between 2 and 5 [20] and increases b. A tokamak with superconducting magnets and A the neutron tolerance by a factor of between 4 and 20. = 2; As mentioned above, the overall tape thickness is nor- c. A compact spherical tokamak with copper mag- mally 100 microns, so if this is the only change made, 30 nets and A=1.5-1.8; the increase in tape thickness will be less than 20%. d. A compact spherical tokamak with LTS magnets [0049] Another approach is to reduce the thickness of and A=1.5-1.8; the copper 505 and hasteloy 501 layers (or other con- e. A compact spherical tokamak with HTS magnets ducting/supporting non-HTS layers in the tape). Halving and A=1.5-1.8; the thickness of these non-HTS layers approximately 35 f. A compact spherical tokamak with a combination doubles the current density in the tape, allowing more of beryllium or aluminium and HTS magnets and space for shielding. A=1.5-1.8; [0050] A third approach is to use a cryogenically cooled beryllium or aluminium central post in the spherical toka- [0055] Options (a) and (b) are well known, but they can mak instead of HTS as shown in figure 6 option B. There 40 only achieve high efficiency if they are large, comparable would be undesirable resistive losses in the beryllium or in size to ITER. Neutral beam injection is important for aluminium, but these can be minimized by cooling, ideally current drive in such devices, but does not make much to 30K or lower and by connecting the beryllium or alu- contribution to neutron or energy production. minium central post to HTS outer arms of the toroidal [0056] Option (c) is limited by the current carrying ca- field coils. Beryllium or aluminium is chosen because it 45 pacity and strength of copper. It is difficult to achieve a has low resistivity at temperatures of 30K or lower and high field on a small device with low aspect ratio because because it is resistant to damage from high energy neu- of the limited space in the centre of the tokamak. In ad- trons. Other elements or materials with these properties, dition the power and cooling requirements for high cur- or similar properties, could also be used. rents through copper are very substantial and this effec- [0051] A fourth means is to use a combination of an 50 tively prevents long pulse operation. outer cryogenically cooled beryllium or aluminium central [0057] Option (d) is viable and the performance of LTS post with an inner part made of HTS as also shown in (materials such as MgB 2, Nb3Sn and NbTi) could be sat- figure 6 option C. The beryllium or aluminium outer pro- isfactory although the design of a low temperature cryo- vides some shielding of the HTS. Cooling, ideally to 30K stat to allow these materials to achieve high field is chal- or lower, and connecting the beryllium or aluminium /HTS 55 lenging. central post to HTS outer arms of the toroidal field coils [0058] Option (e) is better than option (d) because the is still necessary to minimize resistive losses. performance of HTS is better that LTS. HTS can sustain [0052] A combination of these techniques, for example a higher current density at a higher magnetic field and

7 11 EP 2 752 099 B1 12 higher temperature than LTS. This means the size of the [0066] Dnestrovkij et al [9] provided a DINA code sim- HTS conductor and surrounding cryostat can be smaller ulation of the Wilson CTF, and find by using a different than for LTS. In turn this means that the whole device mix of NBI energies (6MW at 40keV and 44MW at can be smaller and cheaper while operating at higher 150keV) they can provide current ramp - up and, aided magnetic field and hence being more efficient. 5 by a larger tritium fraction of 70% (cf 50%) obtain the [0059] Option (f) may be better than option (e) as the same fusion output (50MW) but at considerably lower beryllium or aluminium central post, or partial central plasma current (5.5MA cf 8MA). Although tritium is post, could overcome the problems of damage to the scarce and expensive, the option of using a larger tritium HTS by fast neutrons. fraction to obtain the same neutron output but at lower [0060] The present document focuses on options (e) 10 plasma pressure (and hence improved plasma stability) and (f), an Efficient Compact Fusion Reactor (ECFR) is attractive. based on a compact spherical tokamak with HTS toroidal [0067] Peng et al [10] proposed a larger CTF with field coils, or with a combination of beryllium or aluminium R=1.2m, A=1.5, k=3.07, Bt=1.1-2.2T, Ip=3.4-8.2MA, and HTS toroidal field coils. This provides the opportunity heating power 15-31 MW, bootstrap (self-driven current) to operate small spherical tokamak reactors at relatively 15 fraction -0.5, Q (ratio of fusion power out to input power) high toroidal fields of 3T and above, potentially providing =0.5-2.5, Pfus=7.5-75MW. This CTF also has an option huge increases in the fusion power produced because, of tritium breeding. in the case of thermal fusion, fusion power can increase [0068] Galvao et al [12] studied a ’Multi functional Com- with the fourth power of the toroidal field. In addition, the pact Tokamak Reactor Concept’ a device of major radius energy loss in dissipation in the magnets is greatly re- 20 Ro = 1.2 (some 50% larger than MAST and NSTX), with duced if HTS is used. A=1.6, Ip = 5MA, BTo = 3.5T, and obtained a fusion gain (Q) ∼ 1 for a range of auxiliary heating powers from 5MW Previous studies of ST-based fusion devices to 40MW. Interestingly, at lower powers the maximum Q∼1 gain occurs at ever lower densities, whereas boot- [0061] Before describing the device in detail, it is help- 25 strap current increases almost linearly with density- so ful to consider previous studies of fusion devices based the higher performance options have the advantage of on spherical tokamaks. largest self-driven current. However this study did not [0062] Stambaugh et al [5] in ’The Spherical Tokamak consider the additional neutron production provided by Path to Fusion Power’ described a family of Spherical beam-plasma interactions. Tokamaks (STs) including a Pilot Plant with major radius 30 [0069] More recently, Kotchenreuther et al [11] pro- of R ∼0.7m (plasma current Ip ∼10MA , central toroidal posed a larger Fusion Neutron Source with 100MW fu- field BTo ∼ 2.8T) which have significant output (400MW) sion output (Ro = 1.35m, aspect ratio 1.8, BTo =3.1T, Ip at an optimistically high H-factor (increase in energy con- = 10-14 MA) using their ’Super X’ divertor to solve the finement over scaling law for conventional tokamaks) ∼7 critical divertor thermal load problem. Their device is de- 35 and βT (measure of efficiency: the ratio of plasma pres- signed for use either as a CTF, or as the basis of a fusion- sure contained to magnetic field pressure required)∼ fission hybrid. 62% and a wall loading of 8MW/m2 (wall assumed to be [0070] All the above studies employ NBI for current at radius Ro +2a) and which are designed to produce drive and heating, in conjunction with α-particle heating electricity economically. (note α-particles have low prompt losses at the high plas- [0063] Hender et al [6] considered a Component Test 40 ma currents employed in the above studies). They use Facility (CTF) based on a similarly modest sized ST (R well-understood technology (e.g. copper windings), and ∼0.7m, Ip ∼ 10.3MA, BTo∼3T, fusion output ∼ 40MW at aspect ratios 1.4 ∼ 1.6 (at which sufficient tritium can be a modest H-factor ∼1.3, βN ∼2.6 and wall load (at Ro+2a) bred without need of a centre-column blanket). of ∼0.75MW/m2) designed to produce sufficient neutron [0071] Recently, smaller, lower power compact fusion fluence to test fusion reactor components. 45 neutronsources have beenproposed, with modest fusion [0064] Wilson et al [7] extended the work of Hender et outputs of 1-2MW. Their requirements are significantly al topropose a CTF again ofA ∼1.6,designed toconsume less demanding than those in the above studies, espe- <1kg of tritium per year and specifically to aid the fast- cially the Stambaugh et al study which requires long- track approach to fusion power by testing components pulse operation close to stability limits and at high wall- and materials. Their device has R ∼0.75m, Ip ∼ 8MA, BTo 50 loading to ensure cost-effective electricity production. ∼ 2.8, H-1.3, PNBI ∼ 60MW, and yields Pfus -50MW of Hender and Wilson require high neutron flux for long pe- which about 25% arises from beam-plasma interactions riodsto provide sufficient componenttesting, and operate (discussed further below). at high plasma current. In these recent proposals, de- [0065] Voss et al [8] developed the Wilson design, in- mands on physics limits and on engineering are much creasing the size slightly to R = 0.85m, a = 0.55m, with 55 reduced, however a useful fusion power should be ob- a slight decrease in current and field to 6.5MA and 2.5T, tainable. again assuming H=1.3 , with PNBI = 44MW and Pfus = [0072] Two recent studies are particularly relevant: 35MW.

8 13 EP 2 752 099 B1 14

Kuteev et al [13] specifically addressed the need for ripheral magnetic field lines in a conventional tokamak a small facility developing up to 10MW of fusion pow- 31 and in a spherical tokamak 32. In the conventional er whilst requiring total auxiliary heating and current tokamak 31, magnetic field lines have comparable drive power < 15MW and total power consumption lengths in the region of a favourable curvature (inner, < 30MW. They re-evaluated the smallest (Ro ∼0.5m) 5 high field and stable region) and unfavourable curvature member of the Stambaugh range but under extreme- of magnetic field (outer unstable region). In the spherical ly reduced conditions: Ip∼ 3MA, BTo ∼ 1.35T with a tokamak 32 the field line path in the inner, stable region neutron fluence of ∼ 3 x 10 17 n/s corresponding to a is significantly higher than in the outer, unstable region fusion power of∼ 1 MW and a neutron load and the field line is generally wrapped onto the central 0.1MW/m2.Modelling showsthat neutron production 10 core of the plasma, where the toroidal magnetic field is is more than doubled by the beam-on-plasma effect. highest. As the particle motion in a magnetic trap is bound Importantly for a first pilot device, the build cost was to the field lines, the most straightforward result of an estimated at less than £200M. aspect ratio decrease is an increase in the plasma col- umn magneto-hydrodynamic (MHD) stability. This im- [0073] Thus rather than operating at high plasma cur- 15 proved MHD stability permits either a significant increase rent, it may be possible to employ significant NBI auxiliary in the plasma current, or a decrease in the toroidal mag- heating and enjoy significant neutron production from the netic field strength; this feature has been exploited in the NBI beam-on-plasma interactions noted by Jassby [1]. successful ST experiments, notably START at Culham This effect occurs when injected beams slow down in a [16]. The figure shows the plasma 33 in the START toka- thermal tokamak plasma, and is effective in the ST plas- 20 mak, with sharp plasma edges, demonstrating the excel- mas considered here. lent confinement properties (H-mode) obtainable in an [0074] Sykes et al [2] develop the beam-plasma fusion ST plasma. concept and propose a small spherical tokamak (SCFNS [0078] To date, STs have produced good physics per- = Super Compact FNS) with fusion power (dominated by formance but so far they have low magnetic fields, low beam-plasma fusion) of 1-2MW. SCFNS has parameters 25 heating power and most of them are short pulse devices. R∼0.5m, with Ip=1.5MA (half that of the Kuteev design), The neutron flux is modest as tritium has not been used, and BT = 1.5T. A neutral beam power of only ∼ 6MW is and modelling shows that even if a D-T mix could be sufficient to both maintain the plasma current, and to pro- employed, neutron yield would be small, mainly because vide the fusion power; this low input power reduces the of the low toroidal field. wall and divertor loadings to tolerable values, so that30 [0079] The proposed device Efficient Compact Fusion techniques developed for ITER can be utilised. Reactor (ECFR) is the first ST to have high magnetic [0075] The spherical tokamak represents a low aspect field, high availability, high neutron fluency, low running ratio version of a conventional tokamak and is a crucial costs and the capability to produce net energy over ex- component of the present invention. tended periods of time. [0076] The concept of a spherical tokamak (ST) was 35 first introduced by Jassby [14] and later by Peng [15]. At Main parameters of ECFR the same time, a small low-aspect ratio tokamak GUTTA was constructed and operated at Ioffe Institute, Russia, [0080] TheECFR device is a long-pulse spherical toka- confirming some of unique features of the ST concept. mak with an elongated plasma, and a double-null diver- The first demonstration of the main advantages of a40 tor. The design objectives are to demonstrate routine spherical tokamak (i.e. high beta, high natural elongation, steady-state operation in hydrogen (enabling optimisa- improved stability and enhanced confinement - H-mode) tion and any necessary modifications to be made without was on the START device [16] which was operating at problems of radioactivity), before proceeding to deuteri- CulhamLaboratory 1990-1998.START was a smalltoka- um-deuterium (DD) and then, if desired, to a deuterium- 45 mak but achieved normalised plasma pressures βt ∼ 40% tritium (DT) mix where considerable neutron fluence (which is still a record for tokamaks). In the ST the aspect would result. The design incorporates optional features ratio A of the plasma column is substantially reduced with (notably shielding/neutron reflectors and a heavy water respect to conventional tokamak aspect ratio range blanket) which allow control of the neutron output for test (A≈3-4), giving significant improvements in plasma sta- purposes. bility. The combination of simple construction, excellent 50 [0081] Standard operation produces significant D-T fu- results and high reliability confirmed on more than 15 sion power for a burn length of longer than 1000 sec small and medium sized STs operated during the last which is determined as a "quasi steady-state" for most decade produce a strong motivation for an ST as the next engineering requirements. The injection of neutral step in fusion research, and the high performance and beams of energy up to about 200keV provides the main small size makes the ST economical both in build cost 55 source of auxiliary power and can assist current drive. and in tritium consumption (if D-T operation is desired). RF heating and current drive is also considered. [0077] Figure 3 (courtesy of Y-K M Peng) illustrates an effect of aspect ratio reduction. The figure shows the pe-

9 15 EP 2 752 099 B1 16

Start-up and ramp-up design. [0090] Various methods of heating (and current drive) [0082] In existing tokamaks the plasma current is ini- including NBI and a range of radiofrequency (RF) meth- tiated by transformer action using a large central sole- ods may be appropriate. NBI is the most widely used noid. It is planned to obtain start-up and ramp-up of the 5 scheme and has the advantages of easy injection into plasma current in ECFR without use of a large central the plasma, and less sensitivity to plasma parameters solenoid because, in the final design, the large neutron than most RF methods. fluence may prohibit its use, as there may be insufficient [0091] NBI is also the most commonly used method of space for the extensive shielding required to protect the current drive. Its efficiency depends on many parameters windings. In the present invention a wider range of tech- 10 - beam energy, angle of injection, density of plasma. Typ- niques may be used. ically 1 MW of NBI may drive 0.1 MA of plasma current; [0083] A major advantage of the spherical tokamak is and since NBI costs approx £3M per MW, this is a major that the plasmas (having low aspect ratio and high elon- cost. A potentially helpful feature is the self-driven ’boot- gation) have low inductance, and hence large plasma strap’ current, produced in a hot, high energy, plasma, currents are readily obtained - the input of flux from the 15 which can account for one-half or more of the required increasing vertical field necessary to restrain the plasma current. However bootstrap current increases with den- is also significant at low aspect ratio [17]. sity, whereas NBI current drive reduces at high density, [0084] Experiments on MAST have demonstrated so a careful optimisation is required. start-up using a 28GHz 100kW gyrotron (assisted by ver- tical field ramp) at an efficiency of 0.7A/Watt [18]. A gy- 20 Thermal load on divertors rotron fitted to ECFR could have power∼ 1MW and is predicted to produce a start-up current of ∼700kA. [0092] Some of the energy pumped into a plasma ei- [0085] An alternative scheme is to use a small solenoid ther to heat it or produce current drive emerges along (or pair of upper / lower solenoids) made using mineral the scrape-off-layer (SOL) at the edge of the plasma, insulation with a small shielding (or designed to be re- 25 which is directed by divertor coils to localised divertor tracted before D-T operation begins); it is expected that strike points. The power per unit area here is of critical such a coil would have approximately 25% of the volt- concern in all fusion devices, and would not normally be secs output as an equivalent solenoid as used on MAST acceptable in a small neutron or energy source. However or NSTX. Initial currents of order 0.5MA are expected. in the present proposal the input power is greatly reduced The combination of both schemes would be especially 30 (of order of a few MW, compared to tens of MW in other efficient. designs) so the divertor load is correspondingly reduced. [0086] A novel development of the ’retractable sole- Additional methods are used to reduce the load per unit noid’ concept is to use a solenoid wound from HTS, to area further, by a combination of strike-point sweeping; cool it in a cylinder of liquid nitrogen outside the tokamak, use of the ’natural divertor’ feature observed on START; insert it into the centre tube whilst still superconducting, 35 and use of divertor coils to direct the exhaust plume (as pass the current to produce the initial plasma, then retract advocated by Peng & Hicks [19]); possibly to expand the the solenoid before D-T operation. Advantages of using footprint to large radius as in the ’super - X’ divertor ad- HTS include lower power supply requirements, and the vocated by Kotschenreuther et al [11]. This latter normal- high stresses that can be tolerated by the supported HTS ly requires large currents in the divertor control coils, as winding. 40 these have to be somewhat removed from the neutron [0087] This initial plasma current will be an adequate source for protection: however this demand is made trac- target for the lower energy NBI beams, and the heating table here because of the modest plasma current re- and current drive they produce will provide current ramp quired. Further benefit may be gained by use of a flow up to the working level. of liquid lithium over the target area which will also be 45 used to pump gases from the vessel, for example in a Heating and current drive closed lithium flow loop.

[0088] As previously discussed, it is desirable to obtain General outline of this device a significant fluence of neutrons at minimum auxiliary heating and minimum current drive, in order to minimise 50 [0093] A cross section of a spherical tokamak with con- build costs, running costs, and to keep divertor heat loads ventional copper magnets suitable for use as a neutron at tolerable levels. source is shown in Figure 4. The major components of [0089] Recent energy confinement scalings, derived the tokamak are a toroidal field magnet (TF) 41, optional from recent results on both MAST at CCFE and NSTX small central solenoid (CS) 42 and poloidal field (PF) at Princeton suggest that energy confinement in an ST 55 coils 43 that magnetically confine, shape and control the has a stronger dependence on magnetic field, and a low- plasma inside a toroidal vacuum vessel 44. The centring er dependence on plasma current, than for conventional force acting on the D-shaped TF coils 41 is reacted by tokamaks, and hence is improved for the high field of this these coils by wedging in the vault formed by their straight

10 17 EP 2 752 099 B1 18 sections. The outer parts of the TF coils 41 and external extended well above 1000 s and the system is designed PF coils are optionally protected from neutron flux by a for steady-state operations. The integrated plasma con- blanket(which may be D 2O) and shielding 45. The central trol is provided by the PF system, and the pumping, fuel- part of TF coils, central solenoid and divertor coils are ling (H, D, T, and, if required, He and impurities such as 5 only protected by shielding. N2, Ne and Ar), and heating systems based on feedback [0094] The vacuum vessel 44 may be double-walled, from diagnostic sensors. comprising a honey-comb structure with plasma facing [0098] The pulse can be terminated by reducing the tiles, and directly supported via the lower ports and other power of the auxiliary heating and current drive systems, structures. Integrated with the vessel are optional neu- followed by current ramp-down and plasma termination. tron reflectors 46 that could provide confinement of fast 10 The heating and current drive systems and the cooling neutrons which would provide up to 10-fold multiplication systems are designed for long pulse operation, but the of the neutron flux through ports to the outer blanket pulse duration may be determined by the development where neutrons either can be used for irradiation of tar- of hot spots on the plasma facing components and the gets or other fast neutral applications, or thermalised to rise of impurities in the plasma. low energy to provide a powerful source of slow neutrons. 15 [0099] The approach outlined above enables the de- The reason for such assembly is to avoid interaction and sign of an Efficient Compact Fusion Reactor (ECFR) that capture of slow neutrons in the structures of the tokamak. is much smaller than previous designs of fusion reactors

The outer vessel optionally contains D2O with an option aimed at generating net power, having correspondingly forfuture replacement by other types ofblanket (Pb, salts, lower construction and operational costs (volume from etc.) or inclusion of other elements for different tests and 20 1/5 to 1/15 of existing designs, magnetic field energy and studies. The outer shielding will protect the TF and PF tritium consumption 10 - 100 times lower). The ECFR is coils, and all other outer structures, from the neutron ir- an ideal first device to evaluate previously untested areas radiation. The magnet system (TF, PF) is supported by such as steady-state operation, plasma control, tritium gravity supports, one beneath each TF coil. Ports are operation, etc whilst producing at least 1MW of fusion provided for neutral beam injection 47 and for access 48. 25 neutrons ideal for scientific research, materials tests, pro- [0095] Inside the outer vessel the internal components duction of isotopes for medical and other applications, (and their cooling systems) also absorb radiated heat etc. ECFR is capable of producing net energy over an and neutrons from the plasma and partially protect the extended length of time. As such it can be much more outer structuresand magnet coils from excessiveneutron than a useful demonstration of fusion technology, it can 30 radiation in addition to D2O. The heat deposited in the be the first viable demonstration of a fusion power station. internal components in the vessel is ejected to the envi- [0100] This design is made possible by a novel com- ronment by means of a cooling water system. Special bination of new and established techniques over a wide arrangements are employed to bake and consequently range covering plasma initiation; ramp-up of plasma cur- clean the plasma-facing surfaces inside the vessel by rent; key methods of enhancing neutron production at releasing trapped impurities and fuel gas. 35 relatively low current, field and auxiliary heating; use of [0096] The tokamak fuelling system is designed to in- improved energy confinement; means of varying the neu- ject the fuelling gas or solid pellets of hydrogen, deute- tronenergy in acontrollable and tunablemanner; efficient rium, and tritium, as well as impurities in gaseous or solid means of producing steady-state operation; methods of form. During plasma start-up, low-density gaseous fuel handling the exhaust heat load; special methods of con- is introduced into the vacuum vessel chamber by the gas 40 struction, featuring shielding and optional reflectors to injection system. The plasma progresses from electron- both protect coil windings and control the neutron output; cyclotron-heating and EBW assisted initiation, possibly and the use of HTS to enable exceptionally high toroidal in conjunction with flux from small retractable solenoid(s), fields in a small spherical tokamak. and /or a ’merging-compression’ scheme (as used on [0101] A quarter cross section of a spherical tokamak START and MAST), to an elongated divertor configura- 45 with HTS magnets suitable for use as an energy or neu- tion as the plasma current is ramped up. A major advan- tron source is shown in Figure 6A. The important features tage of the ST concept is that the plasmas have low in- of this tokamak in addition to the major components ductance, and hence large plasma currents are readily shown in figure 4 are a centrepost 61 that can be either obtained if required - input of flux from the increasing HTS or beryllium or aluminium, thermal insulation and vertical field necessary to restrain the plasma being sig- 50 cooling channels 62 to allow the centrepost to be cooled, nificant [18]. Addition of a sequence of plasma rings gen- shielding 63 to prevent neutron damage to the outboard erated by a simple internal large-radius conductor may coil 64 made from HTS, a cryostat 65 to cool the HTS also be employed to ramp up the current. and a vacuum vessel 66 which can be inside or outside [0097] After the current flat top is reached, subsequent the shielding 63. plasma fuelling (gas or pellets) together with additional 55 [0102] There are several options for the centrepost 61. heating leads to a D-T burn with a fusion power in the One option includes HTS with or without neutron shield- MW range. With non-inductive current drive from the ing. Another option is shown in Figure 6B and includes heating systems, the burn duration is envisaged to be an inner part 61a of beryllium, aluminium or another non-

11 19 EP 2 752 099 B1 20

HTS material, a coolant channel 62a, vacuum insulation can increase the neutron yield by a factor of 14 for D-D 62b and thermal insulation 62c. A further option is shown fusion. in Figure 6C and is formed by a combination in which an [0108] Favourable confinement scaling: recent re- inner part 61 b is made of HTS and an outer part 61c is search suggests that energy confinement in an ST has made of beryllium, aluminium or another non-HTS ma- 5 a stronger dependence on magnetic field, and a lower terial that provides some shielding against damage to dependence on plasma current, than the ITER scalings the HTS from neutrons. Additional neutron shielding can derived for conventionaltokamaks. This prediction is very be added to each option, subject to the space constraints promising for the high field, and relatively low plasma in a spherical tokamak. current, of ECFR. [0103] Plasma initiation: methods include merging- 10 [0109] Construction features: insulation of the low- compression; magnetic pumping whereby an oscillating voltage toroidal field coil segments can be by stainless current produces plasma rings which augment the plas- steel which combines great strength and relatively high ma current; use of a retractable solenoid, or pair of such resistance; the TF system may be demountable, utilising solenoids; RF current initiation by a gyrotron. high-duty versions of the feltmetal sliding joints devel- [0104] Current ramp-up: methods include retractable 15 oped by Voss at CCFE; the device itself could feature a solenoid(s), which may be pre-cooled high temperature combination of heavy-water tanks and layers of shielding superconductor solenoids; RF current drive; and the ef- /reflectors (eg of Be or Pb) to protect the PF coils and ficient drive produced by heating the plasma so that the external TF coils from lower energy neutrons, and to di- rapid increase in poloidal field necessary to contain the rect the main stream of neutrons for research and growing plasma inputs almost sufficient flux to ramp up 20 processing tasks. the plasma current to the desired working value. [0110] It is also possible to shoot positive ion beams [0105] Enhanced neutron production: in a conven- directly into the plasma through iron tubes which shield tional fusion device nearly all neutron production arises out the magnetic field. from thermal fusion in the central highest temperature [0111] It will be appreciated that compact fusion reac- region of the plasma. In contrast, in the SCFNS super- 25 tors such as those described herein have a much larger compact neutron source, most neutron production is from surface area per unit plasma volume than bigger toka- interaction of one or more neutral beams with the plasma. maks. In general costs and implementation difficulty In the proposed ECFR device, the high value of the toroi- scale at least linearly with plasma volume, while energy dal field provides high plasma temperatures, and neutron output (which can be considered to be limited by accept- output is a mix of thermal and beam-thermal fusion. New 30 able damage levels) scales linearly with surface area. In modelling shows that neutron production is further en- addition, the costs of a "one (or few) of a kind" device are hanced by the relatively long path of the NBI beam when well known to be higher than the costs of "many of a kind" directed at optimum angle through the highly-elongated devices. It therefore seems likely that many smaller fu- plasma (a natural feature of an ST) and by optimising the sion reactors should be cheaper per unit net power output tritium fraction. The tritium fraction may be optimised by 35 than one large fusion reactor. the use of either deuterium or tritium neutral beams which [0112] It will be appreciated that variations from the will provide refuelling as well as heating and current drive. above described embodiments may still fall within the [0106] Variable neutron energy: in a conventional fu- scope of the invention, as defined in the appended sion device the neutron energy is fixed at 14 MeV for D- claims. T fusion and 2.5 MeV for D-D fusion. In one version of 40 the proposed device an antenna configured to induce ion References cyclotron resonance heating (ICRH) would be mounted inside the toroidal chamber. This ICRH system could also [0113] be configured to increase the energy of the emitted neu- trons by several MeV in a controllable and tunable man- 45 [1] Jassby D L ’Optimisation of Fusion Power Density ner. in the Two Energy Component Tokamak Reactor’ [0107] Optimising neutron output from D-D fusion: Nuclear Fusion 1975 Vol 15 p453 while D-T fusion is the best way to achieve the highest [2] A. Sykes et al, ’Fusion for Neutrons - a realizable neutron flux and energy for some applications, it may be fusion neutron source’, Proc of 24th IEEE Symposi- more effective to avoid the problems associated with tri- 50 um on Fusion Enginering, Chicago 2011 Invited Pa- tium (eg cost, complexity, safety, regulation or availabil- per SO2B-1 ity) and instead use ICRH to increase neutron energy [3] M.Valovic et al, Nuclear Fusion 51 (2011) 073045 and/or to heat a D-D plasma to increase neutron flux. [4] C Hellesen et al, Nuclear Fusion 50 (2010) This use of ICRH can be combined with higher toroidal 022001 field and higher plasma current to give a surprisingly high 55 [5] R D Stambaugh et al, ’The Spherical Tokamak neutron output from D-D Fusion in a system that may be Path to Fusion Power’, Fusion Technology Vol 33 more cost effective than a D-T Fusion system producing P1 (1998) the same neutron flux. Data from JET [4] shows that ICRH [6] T C Hender et al, ’Spherical Tokamak Volume

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Neutron Source’, Fusion Engineering and Design 45 that the major radius of the confined plasma is (1990) p265 1.5 m or less. [7] H R Wilson et al ’The Physics Basis of a Spherical Tokamak Component Test Facility Proc. 31st EPS 2. The fusion reactor of claim 1, wherein the plasma Conf 2004 5 includes tritium ions. [8] G M Voss et al, ’Conceptual design of a Compo- nent Test Facility based on the Spherical Tokamak’, 3. The fusion reactor of claim 1, wherein the plasma FED 83 (2008) p1648 includes deuterium ions but not tritium ions. [9] A Dnestrovskij et al, Plasma Devices and Oper- ations, 15, 2007, p1 10 4. The fusion reactor of any preceding claim, further [10] Y-K M Peng et al 2005 Plasma Phys. Control. including one or more of the following features: Fusion 47 B263 [11] M. Kotschenreuther, P. Valanju, S. Mahajan, neutral beams are directed into the plasma from L.J. Zheng, L.D. Pearlstein, R.H. Bulmer, J. Canik different directions selected to optimise fusion and R. Maingi ’The super X divertor (SXD) and a15 reactions between particles in the beams; compact fusion neutron source (CFNS)’ Nucl. Fu- the reactor is configured so that power input to sion 50 (2010) 035003 (8pp) the plasma is less than 100 MW, preferably less [12] R M O Galvao et al, ’Physics and Engineering than 10 MW, more preferably less than 6 MW, Basis of a Multi-functional Compact Tokamak Reac- more preferably less than 3 MW, more prefera- tor Concept’, paper FT/P3-20, IAEA conf 2008 20 bly less than 1 MW; [13] B V Kuteev et al, ’Plasma and Current Drive the reactor is arranged to operate as a neutron parameter options for a low-power Fusion Neutron source or an energy source; and Source’ 23rd IEEE/NPSS Symposium on Fusion En- the plasma current is driven without induction. gineering, 2009. SOFE 2009. [14] DL Jassby, Comments PlasmaPhys. Controlled 25 5. The fusion reactor of any preceding claim in which Fusion, 3 (1978) 151 the material from which the toroidal field magnets [15] Y-K.M. Peng and D.J. Strickler, Nucl. Fusion 26, are constructed includes an HTS layer within the ma- 769 (1986). terial having a thickness of greater than about 1 mi- [16] M Gryaznevich at all, Phys. Rev. Letters, 80, cron, preferably greater than 2 microns. (1998) 3972 30 [17] O. Mitarai and Y. Takase, Plasma current ramp- 6. The fusion reactor of any preceding claim in which up by the outer vertical field coils in a spherical toka- the material from which the toroidal field magnets mak reactor, Fusion Sci. Technol. 43 (2003), are constructed includes non-HTS layers having a [18] V. Shevchenko, Nuclear Fusion Vol 50 (2010) combined thickness of less than about 90 microns p22004 35 or HTS layers having a thickness greater than about [19] Y-K M Peng and J B Hicks,: proceedings of the 1 micron within the HTS manufactured material. 16th Symposium on Fusion Technology, London, U.K., 3-7 September 1990, Vol 2 p1288 7. The fusion reactor of any preceding claim comprising [20] V. Selvamanickam, "2G HTS Wire for High Mag- a central column formed from non-HTS material in- netic Field Applications," HTS4Fusion Conductor 40 cluding beryllium or aluminium. Workshop, May 26-27, 2011, Karlsruhe, Germany. 8. The fusion reactor of any of claims 1 to 6, wherein an inner part of a central column is made of HTS and Claims an outer part is made of a non-HTS material com- 45 prising beryllium or aluminium that provides shield- 1. A compact spherical tokamak nuclear fusion reactor ing against damage to the HTS from neutrons. comprising a toroidal plasma chamber and a plasma confinement system arranged to generate a mag- 9. The fusion reactor of any preceding claim, wherein netic field for confining a plasma in the plasma cham- the non-HTS material including beryllium or alumin- ber, wherein: 50 ium is cryogenically cooled to reduce its resistance and is optionally joined to HTS material forming the the plasma confinement system includes toroi- remainder of the toroidal field magnets apart from dal field magnets (41) made from material com- the central column. prising high temperature superconductor (HTS) cooled in use to 30K or less and configured such 55 10. The fusion reactor of any preceding claim, further that the magnetic field in use includes a toroidal comprising divertors optimised to reduce the load component of 5T or more, thereby allowing the per unit area on the walls of the plasma chamber, plasma confinement system to be configured so wherein part of all of the surface of the divertors is

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optionally coated with lithium. 4. Fusionsreaktor nach einem der vorstehenden An- sprüche, ferner einschließend eines oder mehrere 11. The fusion reactor of any preceding claim wherein der folgenden Merkmale: part or all of the surface of the plasma facing wall is coated with lithium. 5 neutrale Strahlen werden aus verschiedenen Richtungenin das Plasma gerichtet, ausgewählt 12. A power station comprising a plurality of fusion re- zum Optimieren von Fusionsreaktionen zwi- actors as claimed in any preceding claim. schen Teilchen in den Strahlen; der Reaktor ist so konfiguriert, dass die in das 13. A method of generating neutrons or energy by op- 10 Plasma eingegebene Leistung weniger als 100 erating a spherical tokamak nuclear fusion reactor MW beträgt, vorzugsweise weniger als 10 MW, comprising a toroidal plasma chamber, the method stärker bevorzugt weniger als 6 MW, stärker be- comprising: vorzugt weniger als 3 MW, stärker bevorzugt weniger als 1 MW; initiating a plasma in the plasma chamber; 15 der Reaktor ist angeordnet, als eine Neutronen- using toroidal field coils comprising high temper- quelle oder eine Energiequelle zu arbeiten; und ature superconductor material cooled to 30K or der Plasmastrom wird ohne Induktion angetrie- less to generate a magnetic field with a toroidal ben. component of 5T or more to confine the plasma in the plasma chamber, the plasma having a ma- 20 5. Fusionsreaktor nach einem der vorstehenden An- jor radius of 1.5 m or less; and sprüche, wobei das Material, aus dem die ringförmi- emitting neutrons and other energetic particles. gen Feldmagneten gebaut sind, eine HTS-Schicht innerhalb des Materials mit einer Dicke von mehr als 14. The method of claim 11 or 12, further comprising etwa 1 Mikrometer, vorzugsweise mehr als 2 Mikro- maintaining the plasma in a steady state for at least 25 meter, einschließt. 10000 seconds. 6. Fusionsreaktor nach einem der vorstehenden An- 15. The method of any of claims 11 to 13, wherein the sprüche, wobei das Material, aus dem die ringförmi- neutrons are generated at a rate of at least 3 x 10 17 gen Feldmagneten gebaut sind, Nicht-HTS-Schich- neutrons per second. 30 ten mit einer kombinierten Dicke von weniger als et- wa 90 Mikrometer oder HTS-Schichten mit einer Di- cke von mehr als etwa 1 Mikrometer innerhalb des Patentansprüche HTS-hergestellten Materials einschließt.

1. Kompakter kugelförmiger Tokamak-Kernfusionsre- 35 7. Fusionsreaktor nach einem der vorstehenden An- aktor, umfassend eine ringförmige Plasmakammer sprüche, umfassend eine zentrale Säule, gebildet und ein Plasmaeinschlusssystem, angeordnet zum aus Nicht-HTS-Material, einschließend Beryllium Erzeugen eines Magnetfelds zum Einschließen ei- oder Aluminium. nes Plasmas in der Plasmakammer, wobei: 40 8. Fusionsreaktor nach einem der Ansprüche 1 bis 6, das Plasmaeinschlusssystem ringförmige Feld- wobei ein innerer Teil einer zentralen Säule aus HTS magnete (41) einschließt, hergestellt aus einem hergestellt ist und ein äußerer Teil aus einem Nicht- Material, umfassendeinen Hochtemperatur-Su- HTS-Material hergestellt ist, enthaltend Beryllium praleiter (HTS), im Gebrauch auf 30 K oder we- oder Aluminium, das Abschirmung gegen Schäden niger gekühlt, und derart konfiguriert, dass das 45 an dem HTS durch Neutronen bereitstellt. Magnetfeld im Gebrauch eine ringförmige Kom- ponente von 5 T oder mehr einschließt, wodurch 9. Fusionsreaktor nach einem der vorstehenden An- gestattet wird, das Plasmaeinschlusssystem so sprüche, wobei das Nicht-HTS-Material, das Beryl- zu konfigurieren, dass der Hauptradius des ein- lium oder Aluminium einschließt, kryogen gekühlt geschlossenen Plasmas 1,5 m oder weniger be- 50 wird, um seinen Widerstand zu reduzieren, und trägt. wahlweise mit einem HTS-Material, das den Rest der ringförmigen Feldmagnete abgesehen von der 2. Fusionsreaktor nach Anspruch 1, wobei das Plasma zentralen Säule bildet, verbunden ist. Tritiumionen einschließt. 55 10. Fusionsreaktor nach einem der vorstehenden An- 3. Fusionsreaktor nach Anspruch 1, wobei das Plasma sprüche, ferner Divertoren umfassend, die optimiert Deuteriumionen einschließt, aber keine Tritiumio- sind, um die Last pro Einheitsfläche an den Wänden nen. der Plasmakammer zu reduzieren, wobei ein Teil

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oder die Gesamtheit der Oberfläche der Divertoren 2. Réacteur de fusion selon la revendication 1, dans wahlweise mit Lithium beschichtet ist. lequel le plasma inclut des ions de tritium.

11. Fusionsreaktor nach einem der vorstehenden An- 3. Réacteur de fusion selon la revendication 1, dans sprüche, wobei ein Teil oder die Gesamtheit der 5 lequel le plasma inclut des ions de deutérium mais Oberfläche der zum Plasma weisenden Wand mit pas d’ions de tritium. Lithium beschichtet ist. 4. Réacteur de fusion selon l’une quelconque des re- 12. Kraftwerk, umfassend eine Vielzahl von Fusionsre- vendications précédentes, incluant en outre une ou aktoren nach einem der vorstehenden Ansprüche. 10 plusieurs des caractéristiques qui suivent :

13. Verfahren zum Erzeugen von Neutronen oder Ener- des faisceaux de neutres sont dirigés à l’inté- gie durch Betreiben eines kugelförmigen Tokamak- rieur du plasma depuis des directions différen- Kernfusionsreaktors, umfassend eine ringförmige tes qui sont sélectionnées de manière à optimi- Plasmakammer, das Verfahren umfassend: 15 ser des réactions de fusion entre des particules dans les faisceaux ; Initiieren eines Plasmas in der Plasmakammer; le réacteur est configuré de telle sorte que la Verwenden von ringförmigen Feldspulen, um- puissance qui est entrée sur le plasma soit in- fassend ein auf 30 K oder weniger gekühltes férieure à 100 MW, de façon préférable inférieu- Hochtemperatur-Supraleiter-Material zum Er- 20 re à 10 MW, de façon davantage préférable in- zeugen eines Magnetfelds mit einer ringförmi- férieure à 6 MW, de façon davantage préférable gen Komponente von 5 T oder mehr zum Ein- inférieure à 3 MW, de façon davantage préféra- schließen des Plasmas in der Plasmakammer, ble inférieure à 1 MW ; wobei das Plasma einen Hauptradius von 1,5 m le réacteur est agencé de manière à fonctionner oder weniger aufweist; und 25 en tant que source de neutrons ou en tant que Emittieren von Neutronen oder anderer energie- source d’énergie ; et reicher Teilchen. le courant de plasma est produit sans induction.

14. Verfahren nach Anspruch 11 oder 12, das ferner den 5. Réacteur de fusion selon l’une quelconque des re- Schritt umfasst, dass das Plasma mindestens 10000 30 vendications précédentes, dans lequel le matériau Sekunden in einem stationären Zustand gehalten à partir duquel les aimants de champ toroïdaux sont wird. construits inclut une couche en HTS à l’intérieur du matériau qui présente une épaisseur supérieure à 15. Verfahren nach einem der Ansprüche 11 bis 13, wo- environ 1 micromètre, de façon préférable supérieu- bei die Neutronen bei einer Rate von mindestens 3 35 re à 2 micromètres. x 1017 Neutronen pro Sekunde erzeugt werden. 6. Réacteur de fusion selon l’une quelconque des re- vendications précédentes, dans lequel le matériau Revendications à partir duquel les aimants de champ toroïdaux sont 40 construits inclut des couches non en HTS qui pré- 1. Réacteur de fusion nucléaire tokamak sphérique sentent une épaisseur combinée inférieure à environ compact comprenant une chambre à plasma toroï- 90 micromètres ou des couches en HTS qui présen- dale et un système de confinement de plasma qui tent une épaisseur supérieure à environ 1 micromè- est agencé de manière à générer un champ magné- tre à l’intérieur du matériau fabriqué en HTS. tique pour confiner un plasma dans la chambre à 45 plasma, dans lequel : 7. Réacteur de fusion selon l’une quelconque des re- vendications précédentes, comprenant une colonne le système de confinement de plasma inclut des centrale qui est formée à partir d’un matériau non aimants de champ toroïdaux (41) réalisés à par- HTS qui inclut du béryllium ou de l’aluminium. tir d’un matériau comprenant un supraconduc- 50 teur haute température (HTS) qui est refroidi en 8. Réacteur de fusion selon l’une quelconque des re- utilisation jusqu’à 30 K ou moins et configurés vendications 1 à 6, dans lequel une partie interne de telle sorte que le champ magnétique en uti- d’une colonne centrale est réalisée en HTS et une lisation inclue une composante toroïdale de 5T partie externe est réalisée en un matériau non HTS ou plus, d’où ainsi la possibilité que le système 55 qui comprend du béryllium ou de l’aluminium qui as- de confinement de plasma soit configuré de telle sure une protection vis-à-vis d’un endommagement sorte que le rayon principal du plasma confiné au niveau du HTS du fait de neutrons. soit de 1,5 m ou moins.

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9. Réacteur de fusion selon l’une quelconque des re- vendications précédentes, dans lequel le matériau non HTS qui inclut du béryllium ou de l’aluminium est refroidi de façon cryogénique de manière à ré- duire sa résistance et est en option joint à un maté- 5 riau HTS qui forme le reste des aimants de champ toroïdaux indépendamment de la colonne centrale.

10. Réacteur de fusion selon l’une quelconque des re- vendications précédentes, comprenant en outre des 10 diverteurs qui sont optimisés de manière à réduire la charge par aire unitaire sur les parois de la cham- bre à plasma, dans lequel une partie ou la totalité de la surface des diverteurs est en option revêtue de lithium. 15

11. Réacteur de fusion selon l’une quelconque des re- vendications précédentes, dans lequel une partie ou la totalité de la surface de la paroi faisant face au plasma est revêtue de lithium. 20

12. Centrale électrique comprenant une pluralité de réacteurs de fusion comme revendiqué selon l’une quelconque des revendications précédentes. 25 13. Procédé de génération de neutrons ou d’énergie en faisant fonctionner un réacteur de fusion nucléaire tokamak sphérique comprenant une chambre à plasma toroïdale, le procédé comprenant : 30 l’initiation d’un plasma dans la chambre à plasma ; l’utilisation de bobines de champ toroïdales comprenant un matériau supraconducteur hau- te température qui est refroidi jusqu’à 30 K ou 35 moins pour générer un champ magnétique qui présente une composante toroïdale de 5T ou plus afin de confiner le plasma dans la chambre à plasma, le plasma présentant un rayon prin- cipal de 1,5 m ou moins ; et 40 l’émission de neutrons et d’autres particules énergétiques.

14. Procédé selon la revendication 11 ou 12, compre- nant en outre le maintien du plasma dans un état 45 stable pendant au moins 10 000 secondes.

15. Procédé selon l’une quelconque des revendications 11 à 13, dans lequel les neutrons sont générés à un débit d’au moins 3 x 1017 neutrons par seconde. 50

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REFERENCES CITED IN THE DESCRIPTION

This list of references cited by the applicant is for the reader’s convenience only. It does not form part of the European patent document. Even though great care has been taken in compiling the references, errors or omissions cannot be excluded and the EPO disclaims all liability in this regard.

Non-patent literature cited in the description

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