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Ca9110926 ALTERNATE FUSION CONCEPTS

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Canadian Fusion Fuels Technology Project

ca9110926

ALTERNATE FUSION CONCEPTS: STATUS AND PLANS

CFFTP-G-9009 October 1990

P.J. Gierszewski, A.A. Harms* and S.B. Nickerson' ALTERNATE FUSION CONCEPTS: STATUS AND PLANS

CFFTP-G-9009 October 1990

P.J. Gierszewski, A.A. Harms* and S.B. Nickerson'

McMaster University Ontario Hydro Research Division CFFTP-G-9009

Prepared by: P.J. GierszewskiO Fusion Systems Engineer Fuel Systems & Materials Development Canadian Fusion Fuels Technology Project

Reviewed by:

Manager Fuel Systems & Materials Development Canadian Fusion Fuels Technology Project

Approved by: D.P. Dautovich Program Manager Canadian Fusion Fuels Technology Project ACKNOWLEDGEMENTS

We are grateful to the research groups at Los Alamos National Laboratory CTR Division (HDZP, CPRF, ZT-40, FRX-C, CTX), Spectra Technologies (LSX), Naval Research Laboratory (ZFX), University of Maryland (MS), Oak Ridge National Laboratory (ATF), Imperial College (HZP), Institut Gas lonizzati (RFX) and University of Stuttgart (DPF), who showed us their facilities, clarified the key issues, and discussed their results and program plans. We also particularly wish to thank D. Rej (LANL), A. Robson (NRL), R. Krakowski (LANL), P. Stangeby (UTIAS), J. Linhart (U. Pisa), M. Peng (ORNL) and G. Miley (U. Illinois) who kindly reviewed specific sections of the report. ALTERNATE FUSION CONCEPTS STATUS AND PLANS

Table of Contents

1. Introduction 1

2. Advanced 3

3. 11

4. Spherical Torus 18

5. Reversed-Field 24

6. Dense Z-Pinch 32

7. Field-Reversed Configuration 38

8. 45

9. Ignition Experiments and Reactors 9.1 Ignition 52 9.2 Reactors 53

Appendix A: Other Concepts 62 A.1 () 62 A.2 Electrostatic Confinement 63 A.3 Muon-Catalyzed Fusion 64 A.4 Spherical Pinch 64 A.5 Dense Focus 65 A.6 Linear Systems 66 A.7 Miscellaneous Concepts 67

Appendix B: Inertial Confinement 75 1.

ALTERNATE FUSION CONCEPTS: STATUS AND PLANS

1.0 INTRODUCTION

The focus of the world's fusion program is on tokamaks and . These devices have advanced fusion performance by orders-of-magnitude over the past 20 years, with present large machines poised on the edge of achieving energy breakeven. They are supported by a broad base of experimental machines and theory. And there is a large effort to improve these concepts, such as current-drive, more efficient lasers, and low-activation materials. Reactor studies suggest that these concepts extrapolate to power stations with reasonable energy , cost of electricity, and environmental impact.

Other approaches to fusion have also been proposed to provide the following: avoid perceived problems with conventional fusion as a practical power source (complexity, size, cost, ability to handle advanced fuels); reduce the cost per experiment (e.g., CIT - 300 M$US, LMF - 1000 M$US, and ITER - 5000 M$US capital cost), which limits the ability of national fusion programs to carry out these critical development steps; and broaden the theory and experimental database to support mainline fusion by exploring alternate conditions of physics and engineering.

These alternate approaches have their own technical limits, but as a group suffer relative to tokamaks and lasers from programmatic limits - they have not been funded as long or as well. Therefore, they have a smaller experimental database and need a larger extrapolation to practical fusion. This makes a direct comparison with tokamaks and lasers somewhat subjective. Nonetheless, such judgements are routinely made in program budget decisions. Several reviews have been published which provide a consistent description of the concept status and some degree of comparison [1.1-1.5].

The present review summarizes the status of alternate fusion concepts and their plans for the future. We do not rank the alternate concepts, but rather discuss their status and prospects on a directly comparative framework. The selection of concepts for review itself involves a judgement on feasibility and practicality, but is necessary since so many concepts have been proposed. We adopt the following guidelines: - electromagnetic confinement; - reasonable predictions of net energy gain from pure fusion; - significant recent developments or ongoing international activity.

Therefore, we exclude from detailed review: (1) inertial fusion by choice; (2) muon-fusion since useful energy gain appears to require fusion-fission hybrids; (3) 'cold' fusion as 2. predictions of net energy gain are only speculative; and (4) a variety of magnetic fusion concepts that have not shown enough promise to attract reasonable attention and broad international investment. However, for completeness, most of these latter concepts are summarized briefly in Appendices A and B.

The concepts that are reviewed here are the following: - advanced tokamaks - (including heliotrons, heliacs) - spherical tori - reversed-field pinches - field-reversed configurations - - dense Z-pinches.

Each of these concepts is discussed with respect to the following: basic description of concept, especially novel features relative to tokamak; technical description of their performance, covering all major physics and engineering issues; projections to a reactor; machine parameters as measured and as projected for next-step machines; program pians and needs, especially for use. In the final section, an overall view of the status of each concept with respect to achieving ignition, and with respect to reactor designs is presented.

REFERENCES

1.1 F.F. Chen (ed.), 'Alternate Concepts in Controlled Fusion', Electric Power Research Institute, EPRI ER-429-SR (Palo Alto, CAS May 1977).

.,2 US Congress, Office of Technology Assessment, 'STARPOWER, The US and the International Quest for Fusion Energy', OTA-E-338 (Washington DC, October 1987).

1.3 N.A. Krall, 'Alternate Fusion Concepts as Reactors', in 'Unconventional Approaches to Fusion', B. Brunelli and G. Leotta (eds.), Plenum Press (1982) New York.

1.4 V.E. Haloulakos and R.F. Bourke, 'Fusion Propulsion Study', Air Force Space Technology Center, AL-TR-89-005 (Edwards, CA, July 1989).

1.5 R. Krakowski et al, 'Review of Alternative Concepts for Magnetic Fusion', Proc. 4th Mtg on Tech. of Contr. Nucl. Fusion, King of Prussia, PA, 1980 October 14-17, p.797. 3.

2. ADVANCED TOKAMAKS

INTRODUCTION

The "Tokamak" is a toroidal magnetic confinement concept which uses an externally generated toroidal field and a poloidal field created by an internal induced toroidal plasma current. The resulting helically-twisted lines provide good and confinement properties. The pitch of the helical magnetic field, often called the "safety factor" and defined as the ratio of number of toroidal turns per poloidal turn, must be greater than unity.

The main disadvantages are low 8, pulsed current, and linked magnet coils. The low <3> (< 10%) results from various instabilities and implies a low power density, thereby affecting economics and advanced fuel potential. The plasma current is generally induced by a pulsed ohmic coil, so the machine is pulsed, which leads to engineering fatigue and costs associated with frequent pulse startup and stop. The toroidal, ohmic and poloidal magnet coils are inextricably linked through each other and through the plasma center, which complicates maintenance.

However, the advantages are also well-established. Tokamak theory, experimental database and reactor studies are the most advanced. They are the workhorse of the fusion community and offer the ability to explore reactor-relevant plasma behavior and associated technologies that are useful to most magnetic fusion approaches. And tokamaks continue to improve. Therefore, a review of alternate fusion concepts must consider the tokamak and its potential improvements.

Table 2-1 summarizes the characteristics of representative existing tokamak experiments, and various proposed machines. Figure 2-1 provides a qualitative comparison of present and proposed tokamaks, including ignition and reactor studies.

TECHNICAL DESCRIPTION

Stabilitv/ The tokamak plasma is quite stable, although the limits are constantly being pushed in experiments in order to find better reactor operating conditions. In particular, theoretical and experiment research showed that the usual low 13 limit for tokamaks could be increased. Vertical elongations of up to 2 are commonly used now to improve the beta (e.g., D-lll, JET, CIT, ITER). Early estimates suggested that elongations beyond 2 ("belt pinches") would be even better, but this is not presently considered practical (plasma control becomes difficult [2.11]). Indenting the plasma into a "kidney-bean" shape should also increase the beta limit, as has been shown on PBX. However, this poses considerable engineering difficulties in a reactor. A third approach to higher beta is to use 4. very small aspect ratios (major radius/minor radius, A < 2), the "spherical torus". This case is discussed separately in Section 4.

Within the bounds of "conventional tokamaks", the most promising approach to achieving ~ 10% seems to be to access a second stability regime. Specifically, theory suggests that the 13 limit observed in present tokamaks is primarily due to ballooning instabilities and might be suppressed if S could be made large enough. The first problem is that reaching this high 13 regime seems to require crossing a regime of instabilities that would destroy the plasma in times of 10-100 us. In addition, this B regime may be unstable to kink and resistive instabilities, or could impose impractical constraints on the plasma profiles and transport loss rates [at1]. Proposed approaches to reaching high 6 in tokamaks is by strong plasma shaping, high aspect ratios, and/or shaping the current, pressure and density profiles [2.2,2.8]. Preliminary reactor studies suggest that <6>~10-20%, a high-aspect-ratio circular plasma, a close conducting shell, and low toroidal current and magnetic fields are achievable [2.2,2.10]. Experimentally, the US PBX tokamak is nearing the second-stability regime using a low-aspect-ratio, highly indented bean-shaped plasma, but has so far only confirmed that such shaping can improve plasma behavior [2.8], The US ATF stellarator may have recently reached this regime, which implies that it is stable, although tokamak access would still need to be solved.

Confinement/Transport Confinement and transport are not completely understood, and are usually correlated against several standard scaling laws (e.g., Neo-Alcator, Kaye-Goldston, ITER L-mode). As a representative example of the parametric dependence in an auxiliary heated plasma, 085 12 03 02 05 the ITER-89 power-scaling law for energy confinement is xE ~ I Ro a B (Aj K/P) . Confinement improves with current, field and size, but decreases with auxiliary power. An attractive tokamak reactor demands somewhat better confinement than is predicted by these scaling laws. Operating conditions, called the "H-mode", have been experimentally found in which confinement improves by a factor of two. A major task for present large tokamaks is to understand the reasons for this improvement and to control the plasma in order to take advantage of this mechanism. It is likely that such improvements will come about by more precise control of the microstability of the plasma by adjusting the profiles of current, pressure, density and temperature [2.1]. This would be accomplished by combinations of heating and fuelling systems.

Heating/Current Drive In the original tokamak concept, the plasma current provided both confinement and heating, and was driven inductively by pulsing a transformer. This simple scheme provided excellent performance in the early tokamaks compared to other concepts. However, unless the magnetic field is very high, resistive heating by the plasma current is not sufficient to reach ignition. Also, inductive current drive leads to a pulsed plasma, although the pulses could be for thousands of seconds in reactors. Therefore, tokamaks have been the focus of a significant effort to develop auxiliary heating and current drive 5. methods, based primarily on neutral beams and various radiofrequency waves. The heating technologies have matured quite successfully, although further improvements are still desirable for reactors. Current drive is a relatively new topic (1980's), and present methods are workable but somewhat awkward for reactors (e.g., physically large, inefficient, and/or limited to lower plasma densities). Current drive is a major area of activity for tokamaks.

Impurity Control/Fuelling The critical issue for near-term tokamaks is impurity control. For example, carbon contamination limits present large machines such as TFTR from using their full auxiliary heating system. Dilution of the by low-Z impurities, at typical present levels, reduces the fusion yield by factors of 2-4 and therefore has a major impact on reaching ignition. The design is a critical issue for ITER Technology Phase of operation. Improvements in impurity levels requires better understanding and control of the edge plasma through and pumped limiters, advanced materials for plasma facing components, and possibly novel first wall and divertor designs. Divertor heat removal and lifetime (normal erosion and disruptions) are critical issues for reactors.

Fuelling by gas puffing and pellet injection is sufficient for present machines, but will not be able to directly reach the plasma center in next-step machines like ITER. This may not be critical, but it would be desirable. The behavior of the pellets in plasmas (e.g., under heating), and achieving higher repetition rates, higher velocities and tritium pellets are the major research areas. Present machines largely rely on turbopumps between shots, and surface adsorption during shots. Plasma exhaust for long-pulse machines like ITER requires very large cryopumps or turbopumps, both of which need to be scaled up from present equipment.

Magnets/Power Supplies Tokamak reactors rely on superconducting coils in order to minimize power consumption, in large part due to the low plasma beta. The required coils are under development, and are in use in some machines such as TORE-SUPRA. No particular difficulties are expected for fields up to around 12 T at the coil, but the very large coil sizes are unique to fusion (and tokamaks in particular), and makes the development more expensive.

Since the plasma power density (and even nx) scales roughly as (32B4, then increasing the magnetic field strength may provide a greater economic leverage than increasing 6 [2.5]. While apparently true for any magnetic confinement concept, this may be particularly significant for tokamaks. Higher fields would allow higher 6 and therefore improve the present low power density. High-field tokamaks would also allow larger currents and therefore ohmic heating to ignition, avoiding auxiliary heating and its associated cost, decrease in confinement and uncertainty in confinement scaling.

High-field means 13-25 T at the plasma axis with copper magnets (e.g., ) or 8-13 T with superconductors (e.g., ARIES-I), compared with ~5 T typical for ITER and 6. STARFIRE. These fields can be achieved by various routes: advanced coil design, large aspect ratio, and advanced materials. In the Alcator machines, 14 T was achieved using Bitter-coil magnets with liquid nitrogen cooling. In the Ignitor and Riggatron concepts, the magnets are pushed to high stresses. In Ignitex, single-turn coils are driven with special high-current/low-voltage power supplies. High aspect ratio (A~4-8 rather than 3) tokamaks could allow high fields to be achieved with conventional materials. And new superconducting materials could allow 15-18 T at the coil.

Reactor Engineering Tokamak reactors tend towards large power outputs (-2-4 GWf) and large sizes (R-5-7 m), partly due to physics limits and partly due to economies of scale. Recent design studies have focussed on higher beta (TPSS) and higher field (ARIES-I) in order to improve the effective power density of the fusion core from the 50 kWe/Mg typical of STARFIRE, to values around 100 kWe/Mg. The latter was developed on the basis of general economic arguments, particular by comparison with fission power. wall loads are typically 3-5 MW/m2, and the is extracted by conventional thermal power cycles. The blanket and shielding require development, especially with respect to damage limits, but are fairly conventional in approach. The fusion reactor core is a larger fraction of the overall station cost than in fission power stations. The major safety issues arise from of the blanket and first wall structural materials, the presence of mobile tritium, and the potential for serious damage from current-related plasma disruptions. Relatively frequent component replacement is expected due to plasma and/or neutron damage. This leads to a significant need fcr remote handling equipment, hot cells, and disposal facilities. The development of advanced materials, improved plasma facing component designs, and better plasma control are expected to substantially reduce these concerns in the long-term.

PROGRAM PLANS

JET has recently approached a Q(effective) of 0.8, almost achieving energy breakeven. Dlll-D has recently achieved a peak beta of 10%. Improved confinement is being observed on a variety of tokamaks, leading to a factor of 2 'H-mode' improvement assumed as the ITER reference case.

The next major step for conventional tokamaks involves tritium burning to explore heating and physics. This will first occur in TFTR and JET in the mid-90's. Under serious consideration are short-pulse ignition machines, especially CIT (US) and IGNITOR (Italy), and large engineering test reactors, specifically ITER (International), NET (Europe), and FER ().

In addition to these tritium experiments, other new tokamaks are being considered to extend our understanding of tokamaks and improve impurity control, confinement and beta. These experiments require at least a medium-sized tokamak to have a distinct 7. plasma core region and enough confinement to achieve interesting temperatures. Demonstration of impurity control and current drive over long pulses is seen as particularly important (e.g., STE proposal from GA). Smaller machines have been proposed to study particular ideas such as second-stability-regime access in a high- aspect ratio plasma, and high-field performance. The most significant recent concepts and proposals are summarized in Table 2.1, but none have been approved for construction.

REFERENCES

2.1 F. Ribe and D, Baldwin, 'Fusion Program Planning for the Early to Mid 1990s', Jrnl Fusion Energy, 7(4) (1988) 371.

2.2 G.A. Navratil and T.C. Marshall, "High-Beta Tokamak Operation in the Second Stability Regime", Comments Plasma Phys. Controlled Fusion jH)(4) 1986 185.

2.3 S. Mendelsohn et al, 'SRC - Precursor to Large Scale High Beat Plasma Devices', 1989 IEEE Inter. Conf. on Plasma Science, 1989 May 22-24, Buffalo (paper 6P49).

2.4 R.E. Potok, L Bromberg, D.R. Cohn, 'Energy Transport Requirements for Tokamak Reactors in the Second Ballooning Stability Regime', 1986 IEEE Inter. Conf. on Plasma Science, p.368.

2.5 D.R. Cohn, 'Future Directions in Fusion Research: Super High-Field Tokamaks', Jrnl Fusion Energy, 6_(3) (1987) 281.

2.6 Ignitex Group, Papers on the Ignitex Experiment, 13th Intl. Symp. on Fusion Eng., Knoxville, US, 1989 October 2-6.

2.7 'ITER Conceptual Design - Interim Report', IAEA, Vienna (1989).

2.8 'STARPOWER, The U.S. and the International Quest for Fusion Energy', US Congress, Office of Technology Assessment, October 1987.

2.9 M. Okabayashi et al, 'Initial Results of the PBX-M Experiment',IAEA-CN-50/A-II-2, p.97.

2.10 C.C. Baker et al, Tokamak Power Systems Studies - FY 1985', ANL/FPP/85-2, December 1985.

2.11 M.W. Philips, M.H. Hughes. A.M.M. Todd et al, 'Effect of Shaping on the Equilibrium and Stability of High Current, High Beta Tokamaks', Nice, , 1988 October 12-19, Vol.2 (1988) 65. 8. 2.12 G. Greiger, 'Summaries of the 12th IAEA Inter. Conf. on Plasma Pysics and Controlled , Nice, France, 12-19 October 1988', Nucl. Fusion 29(3) (1989)527.

2.13 R. Albanese et al, 'Scaling Laws and Design Criteria for a Press-Supported Compact Tokamak', Fusion Technology 1988, A.M. Van Ingen et al (eds.), Elsevier Science Publishers (1989), (Proc. 15th SOFT, Utrecht, Netherlands 1988) p.1791.

2.14 JET Team, 'Latest JET Results and Future Prospects', Proc. 12th Conf. Plasma Physics and Controlled Nuclear Fusion Research', Nice, France, 1988 October 12- 19, Vol.1 (1988)41.

2.15 JT-60 Team, Proc. 12th Conf. Plasma Physics and Controlled Nuclear Fusion Research', Nice, France, 1988 October 12-19, Vol.1 (1988) 67.

2.16 FT Team, Proc. 12th Conf. Plasma Physics and Controlled Nuclear Fusion Research', Nice, France, 1988 October 12-19, Vol.1 (1988) 637. 9.

Table 2-1. Characteristics of major tokamak experiments

Facility/Location Year R B. 1 xE T,(0) P« Comments/Status (m) (m) (T) (MA) u(s). (%) (1/m3) (s/m3) (keV) (MW)

Conventional Dill, GA 1982 1.6. 1 2 - - 3.5 8e19 4e18 0.7 - [2.1] ISX-B. Oak Ridge 1985 0.9 0.29 1.8 - - - 1e16 1e15 - - [2.1] TEXT, U. Texas 1986 1.0 0.27 3 0.4 0.3 0.3 5e19 2319 0.7 0.2 [2.11 TEXTOR, Julich 1986 1.75 0.46 2 0.6 3 1 5e19 5t18 0.8 3 PLT. Princeton 1986 1.4 0.4 3.2 0.5 1 0.3 1.5e19 4.5e17 7.1 3 [2.1] DITE, Culham 1986 1.17 0.26 2.7 0.3 0.5 0.9 8e19 2.4eK7 1 2.4 ASDEX, Garching 1988 1.65 0.4 2 0.42 0.5 - 4e19 3e18 1 0.9 TdeV, Montreal 1988 0.86 0.27 1.4 0.25 0.4 0.8 3e19 5e17 0.5 0 JT-60. JAERI 1988 3 0.95 3 3 2 1e20 1e19 2 21 [2.1,2.15) JET, Culham 1988 2.96 17 3.1 3 - - 5e19 6e19 6 7.5 H-mode discharge [2.1,2.14] TFTR, Princeton 1988 2.48 0.85 5 1.4 2 - 2e19 3e18 32 30 [2.1] T-15, USSR 1989 2.4 0.7 1 0.12 - - 3e18 3e17 0.1 - JET, Culham 1990 2.96 1.7 . -7 - - 1e19 1e19 20 30 100 kW fusion, Q~1% STE. US - 2 0.9 5 3.5 4000 <3 1e20 3e19 10 37 Long pulse, 1989 concept [2.11 CIT, US - 2.1 1. 10.4 11 5 5 3.6e20 1.8e20 15 10 Ignited, 1988 proposal [2.12] Ignitor, Italy - 1.17 0.6 12 11 - 6 - - - - Ignited, 1988 proposal [2.12) JIT, UK - 7.5 4.5 4.5 30 - - 7©19 - - - Ignited, 1988 concept [2.12) ITER - 6.0 3.2 4.85 22 400 4.2 1.2e20 4.7e20 22 0 Base case, 1989 concept [2.7,2.12]

Second-stability reqime PBX-M, Princeton 1989 1.6 0.45 1.2 0.5 0.2 6.5 4e19 - - 2 SRX, US - 1.5 0.6 1.3 0.16 - - - - - 3 1989 concept [2.1,2.3]

High-field Alcator-C, MIT 1984 0.64 0.165 14 - - 1.5 1.5e21 8e19 1.6 - [2.1.2.51 FT. Italy 1988 0.83 0.2 8 0.36 - - 1e20 3.5e18 2 0.2 [2.16] Ignitex, U. Texas - 1.50 0.61 20.2 12 5 0.6 3.6e20 2e20 12 - Ignited, 1989 proposed [2.6] Ignited SHOT, MIT - 1.95 0.42 18 5.0 5 - - - - - Ignited, 1989 concept [2.5] OMITRON, Italy - 0.4 0.2 17.5 5.0 ------Ignited, 1988 concept [2.13]

* Simultaneous values (some inferred). is average major radius defined as (a+b)/2 for non-circular plasmas. A date is given for achieved results, none if values are projected. ITER 1000 MWf

\

ARIES-I i 1000 MWe

\ "SfARFIRE V 1200

JT-)60 i 8 10 MAJOR RADIUS (m)

Figure 2-1. Major characteristics of several present and proposed tokamaks Solid shapes are actual experiments. Plasma shapes are only indicative. 11.

3. STELLARATOR

INTRODUCTION

A Stellarator is a toroidal confinement concept similar to a tokamak except that the poloidai fields are produced by external magnets rather than a toroidal plasma current. The key consequences of the generation of poloidal field by external magnets are the following: (1) steady-state operation without current drive; (2) currentless operation with no current-driven major disruptions; and (3) plasma startup on existing magnetic surfaces.

The major disadvantage is that stellarators appear to have low beta limits (comparable to tokamaks) and a larger aspect ratio (than tokamaks). They are also non-axisymmetric, which complicates both machine design and theoretical analysis of experimental results.

Unlike a tokamak, the poloidal field, plasma current and toroidal field can be independently varied in a stellarator. This has given rise to a variety of possible magnetic field configurations. The major field characteristics are its rotational transform, shear and well. Transform refers to the twist of the field lines, and is the inverse of the tokamak "safety factor". Shear is the gradient in the transform as one goes radially out from the center of the plasma. A magnetic-well occurs when the magnetic field strength increases radially. Four different design approaches are under consideration: [3.1] - higfvtransform, high-shear (e.g., Heliotron-E) - moderate-transform, moderate-shear, moderate magnetic-well (e.g., ATF) - moderate-transform, low-shear, magnetic-well (e.g., W VII-AS) - high-transform, low-shear, magnetic-well (e.g., TJ-II). Each approach represents a different compromise on plasma and engineering properties, and it is not yet clear which is best.

Furthermore, a given magnetic field configuration can be achieved by several magnet coil configurations. These more visible machine characteristics are often used to classify the machine type (see Figure 3-1): - "classical" stellarator (e.g., Wendelstein Vll-A, L-2, Proto-Cleo) - heliotron/torsatron (e.g., Heliotron-E, ATF) - modular stellarator (e.g., Wendelstein VII-AS) - helical-axis stellarator (e.g., Figure-8 Stellarator, TJ-II, H-1).

In the "Classical" Stellarator. the toroidal field is generated by toroidal field coils, and the poloidal field is generated by pairs of helical windings wrapped toroidally around the plasma, with the current flowing in opposite direction in each half of a winding pair. The advantage is that the poloidal and toroidal fields are independently controllable. The disadvantages are the interlinked coil sets, limited plasma access, and high magnetic forces on the helical windings.

In the Heliotron/Torsatron. the toroidal and poloidal fields are generated by the same helical windings, where the current flows in the same direction in all windings. Separate toroidal field 12. coils are not needed, but there is less flexibility in operation.

In the Helical-Axis Stellarator, the plasma twist is achieved by making the plasma axis itself helical. There are several ways to do this, but the preferred Heliac places toroidal field coils on a helical axis around a central circular toroidal coil (which in turn has a single helical winding around it). In principle, it is possible to avoid the interlinked toroidal coil by use of "bean"-shaped plasmas and/or warped toroidal field coils.

In order to avoid the maintenance difficulties of helical toroidal windings, Modular-coil Stellarators were proposed in which the helical windings were replaced by separate, warped poloidal coils. This improves engineering and has some field profile advantages, but generally results in lower transform, shear and other magnetic field characteristics.

Table 3-1 summarizes the major completed, existing, and planned stellarator experiments.

TECHNICAL DESCRIPTION

Stabilitv/Beta In the 1970's, stellarators were thought to be limited to low B. However, dramatic improvements in theory and experiments in the early 1980's changed the picture. The four current-less magnetic configurations under study are believed capable of achieving ~5- 10% at moderate (A~5) to high (A>10) aspect ratio [3.1,3.2,3.5]. In some cases, these 6 limits require accessing the second stability regime. This is easier in stellarators than in tokamaks, and appears to have been demonstrated recently in ATF. These values of 6 are comparable to those predicted for improved tokamaks. However, present stellarator experiments cannot achieve these 8 values as designed. Verifying and improving these beta limits is a task for future machines, but in general there is confidence that such values are achievable.

Confinement/Transport Good confinement in stellarators requires reducing magnetic islands by careful field design, and reducing plasma current by using auxiliary rather than ohmic heating [3.1]. heat conduction is neoclassical but the edge electron heat conduction (which dominates overall 024 119 131 41 transport) is anomalous [3.2]. Scaling of TE[S] = 0.065 P[MWr BJT] a[m] R0[m]° has been suggested ("LHD" scaling plus density limit) [3.7]. However, the range of explored is still small, and in particular tests with collisionless plasmas with realistic edge conditions are needed. Overall, confinement is similar to tokamaks. Some increase over neoclassical transport is desirable, possibly through magnetic field shaping, electric fields, or edge control (e.g., as in the tokamak H-mode) [3.7].

Heating/Current Drive Early stellarators relied on a toroidal plasma current in order to initiate and heat the plasma. With auxiliary heating in the 70's, net-current-free operation was achieved [3.1], and the pulse length of experiments was limited by design constraints on power supplies and cooling. The magnitude of the bootstrap current remains of interest as it can significantly affect the magnetic field design for reactors. Neutral beam and electron cyclotron heating have been found to be equally efficient on stellarators as on tokamaks [3.6], ion cyclotron heating has 13. been less explored to date.

Impurity Control/Fuelling Impurity control is expected to be a significant issue, comparable to tokamaks. Particle exhaust can in principle be achieved through the natural divertor geometry of heliotron/torsatrons, or by pumped limiters placed in magnetic islands around the outer flux surface. Particle removal has not been well-studied in experiments to date; there are signs of impurity accumulation under conditions which otherwise give good energy confinement [3.5]; and it is more difficult than in tokamaks to increase the plasma/wall separation since this increases the distance from the external coils [3.7]. Particle and impurity removal is a key issue for stellarator research. This will require larger plasma radii for improved impurity screening, and longer pulses to demonstrate steady-state conditions. Fuelling should be comparable to tokamaks; pellets have been used successfully on Heliotron-E and W-VII-A.

Magnet/Power Supplies Magnet design and fabrication is a key engineering difference between tokamaks and stellarators. Non-circular, non-planar coils must be built and joined with high precision. Copper- helical and non-planar modular coils have been built successfully with 1-mm accuracies over 4 and 0.5 m diameters, respectively [3.2]. Experience with W-VIIAS shows that although the non-planar coils take more fabrication, there are no fundamental problems. Superconducting helical coils require superconducting joints, and are under development. Superconducting modular coils should be able to apply the copper modular coil technology, but has not been demonstrated.

Reactor Engineering Stellarator reactor designs show lower neutron wall loads (1.3 to 4 MW/m2 [3.2]) compared with tokamak studies, but overall the fusion core mass power density seems comparable. The larger aspect ratio does make access, and thus maintenance, simpler. Nuclear aspects are generally similar to tokamaks. Stellarators are inherently steady-state, which simplifies engineering and increases reliability. The lack of plasma current is a major advantage for stellarators as it eliminates current-driven disruptions. Stellarator reactors are relatively large, high power (3-4 GWf) machines, with similar balance of plant to tokamaks.

FUTURE PLANS

In 1980-85, advances in theory and computational methods led to accurate treatment of the 3-D nature of stellarators and showed that <8>~5-10% was achievable, thus significantly improving reactor prospects. Experimentally, high-power neutral beam and RF heating systems replaced the ohmic heating current, leading to current-less operation and substantially improved plasmas. And on the design front, new coil concepts and design tools were developed. As a result, the period 1985-90 saw the design, construction and commissioning of major new experiments (ATF, US and Wendelstein-VII/AS, FRG), construction of smaller- scale heliacs (H-1, Australia and TJ-II, Spain), and the design and approval to build a new, very large stellarator (LHD, Japan). 14. The present generation of current-less machines in operation and under construction should determine the best field configuration with respect to confinement, beta limits and operation control under finite-beta, less collisional, and higher auxiliary heating conditions.

Figure 3-2 provides some perspective on the relative size and magnetic field requirement between existing and projected stellarators, including some conceptual reactor designs. The next-generation of stellarators must provide confinement data in more reactor-relevant plasmas with realistic edge conditions (long-pulse, active particle removal) and superconducting coils. Both LHD and W-VII-X would be appropriate size machines. In addition to this size scaling, a unique experiment could explore the low-aspect ratio limit. Presently, stellarators are larger than tokamaks, but there is some belief that the behavior at low aspect ratios (e.g., decrease in 3) may be acceptable and would allow much smaller and less expensive stellarator reactors. A moderate-sized low-aspect ratio torsatron might be built as a ~100 M$ machine, as per the ATF-II conceptual design. A DT experiment could follow these next-step machines, but there is as yet no serious ignition stellarator design.

REFERENCES

3.1 B.A. Carreras et al, 'Progress in Stellarator/Heliotron Research: 1981-1986', Nucl Fusion 28(9) (1988) 1613.

3.2 B.A. Carreras et al, 'Progress in Stellarator/Heliotron Research: 1981-1986, Executive Summary', ORNL/TM-10482, Sept 1987.

3.3 K. Uo et al, 'Recent Results on Heliotron-E1, Nucl. Inst. & Methods, 207 (1983) 151.

3.4 R. Behrisch, O. Gruber, J.-M. Noterdaeme, F. Rau, 'Report on the 14th European Conf. on Contr. Fusion and Plasma Phys., Nucl. Fusion, 27(9) (1987) 1557.

3.5 J. Sheffield, 'Alternative Concepts', Nucl. Fusion, 29(3) (1989) 499.

3.6 L.M. Kovrizhnykh, 'Heating and Confinement of a Stellarator Plasma - A Review of Theory and Experiments (1980-1985)', Plasma Phys. and Contr. Fusion, 30(2) (1988) 67,

3.7 J.F. Lyon (ed.), Special Issue on Stellarators, Fusion Tech. 17(1) (1990).

3.8 S. Yoshikawa and T.H. Stix, 'Experiments on the Model-C Stellarator', Nucl. Fusion 25(9) (1985) 1275. 15.

Table 3-1. Major stellarator experiments*

Facility/Location Year Ft Bo 1 tE T,(0) Comments/Status 3 3 (m) (m) (T) (MA) (s) (%) (1/m ) (s/m ) (keV) (MW)

Present/completed Model-C, PPPL 1969 0.2 0.06 3.5 0.002 0.005 - 2e18 2e14 0.4 4 Racetrack [3.8] CLEO, Culham 1981 0.9 0.13 - 0.05 0.005 10 2e18 2e15 0.06 0.003 W-VII-A, Garching 1986 2 0.1 3.2 -0 0.5 0.45 1.2e20 2.5e18 0.7 0.46 Classical stellarator [3.7] Heliotron-E, Kyoto 1986 2.2 0.2 1.9 -0 0.5 0.47 1.4e20 5e18 0.3 2.6 Torsatron [3.7] W-VII-AS. Garching 1988 2.0 0.2 1.25 0.002 0.5 - 1e19 - - 0.4 Modular coil ATF, ORNL 1988 2.05 0.27 0.95 0.001 0.05 0.5 2.5e19 3.8e17 0.26 1.1 Torsatron CHS, Nagoya 1988 1 .0 0.2 1.1 0.03 . - 3e18 9e15 0.15 0.4 Torsatron SHATLET-M, Tokyo 1988 0.42 0.05 0.15 ~0 3e-4 4 1e19 - 0.1 - Torsatron, 300 J heating IMS, Wisconsin 1989 0.4 0.04 0.6 . . - - - - - L-2, Moscow 1989 0.11 1.5 0.016 0.008 - 2e19 - 0.1 0.28 Uragan-3M, Kharkov - .0 0.11 2.0 ~0 0.5 1.2 - - - 1.2 Torsatron Uragan-2M, Kharkov - .7 0.22 2.4 -0 - 5 5e19 - 1 5 Torsatron TJ-II, Madrid .5 0.22 1.0 ~0 - . 5e18 2e16 0.17 0.4 Heliac H-1, Canberra - I.O 0.2 1.0 ~0 - 0.1 - - - 0.2 Heliac

Next-step machines LHD. Toki 1 0.5 4 ~0 10 - 1e20 2e19 3 20 Torsatron [3.7] W-VII-X. Garching _ 5.5 0.52 3 ~0 2 1e20 3e19 2 20 Modular coil [3.7] ATF-II, Oak Ridge - 2.0 0.52 4 ~0 1 2e20 4e19 2 10 Torsatron [3.7]

* Simultaneous values (some inferred). Dates are given for experimentally achieved results, no dates are for expected results. "Classical" Stellarator: Heliotron/torsatron: Wendelstein Vll-A Heliotron-E

Modular Stellarator: Heliac: Wendelstein- VIIAS TJ-II

Rgure 3-1. Examples of stellarator configurations [3.1]. /""• [ ASR-A6C \ ATR- 2 "I 4000 MWf I 4000 MWf \ > !> /...

ATF-II LHD I Heliotron-H \ CO \ 3400 MWf i X oModel-C : < i i ! r >W-VII-X I o -*-! ...j... i a Uragan-2M UJ LL Heliotron-E g LU o L-2 < O -VII-AS

?-• i.. i O TJ-II • IMS

©Shatlet-M 0 10 15 20 MAJOR RADIUS (m)

Figure 3-2. Relative size and magnetic field for existing and proposed stellarators. Solid shapes are actual experiments; plasma shapes are based on average radius. 18.

4. SPHERICAL TORUS

INTRODUCTION

A "spherical torus" is a very-low-aspect ratio tokamak. The key advantages are high beta and compact size. The main disadvantage is the high plasma current which must be driven by non-inductive (no transformer) methods.

The spherical torus, like a tokamak, has an externally generated toroidal field and a poloidal field generated by a plasma current. The key difference is that the thickness of the inner core of the torus is made as small as possible by eliminating the inboard blanket, inboard coil shielding and ohmic transformer. All that is left is the toroidal field coil leg, which must be made from copper and inorganic insulators for maximum radiation tolerance, and be easily replaced as it becomes neutron damaged. As a result, the aspect ratio is reduced from ~3 typical for conventional tokamaks to 1.3-1.6, for an overall 'spherical' shape.

This change in the tokamak engineering has important effects on the plasma behavior. For example, the plasma naturally elongates vertically, reducing the high-current poloidal coils as required for elongated tokamaks (like ITER or JET) [4.1]. And secondly, the geometry increases the magnetic field rotational transform, resulting in <(i> ~ 20% or even higher in the second stability regime [4.1].

Finally, independent of its reactor potential, a spherical torus experiment could significantly clarify theories and scaling for conventional tokamaks. In particular it could resolve the aspect ratio scaling of confinement in tokamaks, the key difference between present tokamak scaling laws; and the importance of pressure-driven, intense-heating-driven, and resistive instabilities to the beta limit. [4.1]

The spherical torus emerged as a concept in the 1980's. There has been only one small experiment. Table 4-1 summarizes the characteristics of proposed spherical torus machines. Figure 4-1 illustrates a conceptual ignition experiment design.

TECHNICAL DESCRIPTION

Stabilitv/Beta Good stability is predicted for spherical tori. In particular, the spherical torus geometry increases the rotational transform, allowing higher R than in tokamaks. And the toroidal field of the spherical torus plasma core is also highly paramagnetic, allowing the spherical torus to mimic the high-current/high-beta core of spheromaks while maintaining a tokamak safety factor above unity. Predictions for <6> have ranged from ~20% to as much as 50-70% in the second stability regime.[4.1 ] Although these predictions are based on well-supported tokamak theory, there are no definitive spherical torus experiments to confirm them.

Confinement/Transport Confinement and transport are basically expected to be tokamak-like. The smaller minor 19. radius for given fusion power (due to higher beta) tends to decrease confinement, and indeed some tokamak scaling laws imply poorer ntT for spherical tori [4.3]. On the other hand, there are theoretical suggestions that confinement may be better due to the strong paramagnetism of the spherical torus, or if tE improves with Bp and magnetic shear rather than with BT and R, or if the second stability regime can be accessed [4.1,4.3]. Again, the lack of experiments prevents being conclusive.

Heating/Current Drive The spherical torus relies on a large plasma current - large relative to tokamaks -and this must be supplied by current drive due to the lack of a ohmic transformer coil. Since current drive power scales roughly as (current * major radius), then the spherical torus requires the same current drive power as a tokamak to drive twice the current [4.3]. In general, the spherical torus can make use of the neutral beam or radiofrequency methods under development for tokamaks, but strict reliance on these would be unattractive [4.3]. Instead, it is necessary to rely on newer ideas for helicity injection (which may be easier on a spherical torus) and on significant bootstrap current [4.3]. This is a key issue for spherical tori.

Impurity Control/Fuelling Little information is available on fuelling and exhaust. Ref. [4.3] presents a preliminary DT reactor in which the edge runs hot (3-5 keV at the separatrix), maintained by significant pellet injection fuelling into the plasma periphery and a low-recycle flux from the divertor. The estimated fuelling and pumping rate would be 90 times the ash production rate, about twice that for the ITER tokamak design [4.3,4.4]. As a high power density device, the plasma facing components are expected to be under more stress than in tokamaks, although similar to other high-power-density concepts.

Magnet/Power Supplies Spherical tori simplify the pcloidal shaping coils and eliminate the ohmic transformer solenoid of tokamaks. However, the centra! toroidal-field coil leg must be replaceable, and probably made from high-current-density copper with inorganic insulation. The low-turn designs require high-current, low-voltage dc power supplies such as homopolar generators. [4.3]

Reactor Engineering Nuclear aspects are generally the same as for tokamaks. The lack of inboard blankets reduces breeding but the small aspect ratio makes this effect small (-5%) [4.3]. The machine must be designed for quick replacement of the center copper post. This will result in about 15 Mg (3 m3) of activated copper waste every ~6 months [4.3]. The high plasma currents would make the consequences of disruptions more severe relative to tokamaks, but the stored magnetic energy should be less. The balance-of-plant is similar to tokamaks in general. The lower magnetic field and aspect ratio may make direct convenors practical in the divertor.

FUTURE PLANS

Since introduction in the early 1980's, its theoretical basis has gradually improved. Most recently, it has been under evaluation in the ARIES reactor study. The Heidelberg spheromak 20. was converted into a spherical torus and, although a very small machine with a short lifetime plasma, the results were consistent with a high beta at low aspect ratio.

The key issues are the plasma beta limits and confinement in the low-aspect-ratio regime. The first reasonably definitive experiments would be around 0.1-0.2 MA, 0.2-0.4 m major radius, with an ohmic transformer. The experimental programs would cost about 10 M$US for a green-field site, perhaps 3 M$ for the hardware. These machines would check the scaling laws in the low aspect ratio limit (ie., does a spherical torus have tokamak-like confinement or RFP/spheromak-like confinement). Culham Laboratory has began construction of the START spherical torus, and other experiments are under serious design in Brazil and the USSR. The US is active in theory and design, but has not committed to an experiment. The next step would be a somewhat larger machine with current drive (eliminate the ohmic transformer for a lower aspect ratio and for reactor relevance). The need for significant auxiliary power for current drive and heating (to test beta limits) would add to the experimental cost. A multi-year program is expected to cost around 100 M$; no such experiment is presently planned.

Assuming these next-step experiments confirm acceptable scaling of confinement and beta (particularly second stability regime access) at low aspect ratios, and if current-drive technologies continue to mature, then a tritium-burning machine could be contemplated. For a modest size (R < 1 m) and 8 MA current (< 50 MW current drive power), about 50 MW of fusion power could be produced with 1 -2 MW/m2 neutron wall load. The predicted 3 values for the spherical torus offer a reasonable advanced fuel potential in the first stability regime (comparable to RFP's), excellent capability in the second stability regime. Figure 4-2 illustrates the general size and current needs of the various steps considered for the development of the spherical torus concept.

REFERENCES

4.1 Y-K.M. Peng, 'Remarks', J. Fusion Energy, 8(1/2) (1989) 12.

4.2 Y-K.M. Peng, 'Spherical Torus, Compact Fusion at Low Field', ORNL/FEDC-84/7, February 1985.

4.3 Y-K.M. Peng, 'Initial Characterization of Spherical Torus Reactors for ARIES', FEDC-letter, M-89-ARIES-006, 1989 August 23.

4.4 ITER Team, 'ITER Conceptual Design, Interim Report', 1989 October 30, IAEA.

4.5 T.P. Crowley, 'OLIVE - A Small, Low-Aspect Ratio Tokamak', Abstracts, 1989 IEEE Inter. Conf. on Plasma Science, 1990 May 22-24, Buffalo, New York, paper 2W3.

4.6 H. Bruhns et al, 'Study of the Low Aspect Ratio Limit Tokamak in the Heidelberg Spheromak Experiment', Nucl. Fusion 27(12) (1987) 2178. 21.

Table 4-1. Characteristics of major spherical torus experiments*

Facility/Location Year R Bo I t^, <6> tE Tj(O) Pm, Comments/Status (m) (m) (T) (MA) (s) (%) (1/m3) (s/m3) (keV) (MW)

Proof-of-principle HSE, Heidelberg 1987 0.07 0.07 0.2 0.14 2e-5 20 0.02 - Converted spheromak [4.1,4.6] START, Culham 0.22 0.2 0.15 0.015 - 1990 under construction GLOBUS, Leningrad 0.5 0.4 0.5 0.5 - 1990 approved for construction Proto-ETA, Brazil 0.35 - 0.2 - 1989 under design STX, ORNL 0.45 0.4 0.5 0.9 - 1989 concept [4.1] OLIVE, RPI 0.25 0.22 0.15 0.1 0.01 2 6e18 0.3 - 1989 proposed [4.5]

Reactor scaling Steady-state, ORNL 0.65 0.82 0.5 2 40 10 1989 concept [4.1]

Tritium step Neutron source, ORNL 1.0 1.4 0.6 8 SS 50 50 50 MWf, 1989 concept [4.1] Ignition, ORNL 1.53 1.4 2.0 14.1 100 24 6e19

* Simultaneous values (some inferred). Dates are given for achieved results, no date are for expected values. 0RNL-DWG84342S FED

CENTER / CONDUCTOR

TF RETURN LEGS SHIELD VACUUM VESSEL FIRST WALL VACUUM VESSEL SHIELD FIRST WALL

PF COILS

CENTER CONDUCTOR POST

TF RETURN LEGS

1 2 3 4 5

(a) METERS

Figure 4-1. Spherical torus ignition experiment [4.2]. IQRNL jSJT Reactor !b MWe

10. U \

/ORNL I / Steady State \ / ST I\ HI cc DC D K l\ o IIST ORNX L \ i \ < 1.0

i ! >

• I / QL \ \

v -••'••!

iCulham I 0. 1 2 3 4 MAJOR RADIUS (m)

Figure 4-2. Spherical torus development steps to a reactor. Solid shapes are actual experiments; JET and D-lll are shown for comparison. The plasma shapes are only representative. 24.

5. REVERSED-FIELD PINCH

INTRODUCTION

The Reversed-Field Pinch (RFP) is a toroidal plasma maintained by a weak applied toroidal field and a strong poloidal field produced by a toroidal plasma current. It is similar to the tokamak in that magnet coils go through the center of the torus, and a toroidal current is needed. Unlike a tokamak, the toroidal field within the plasma is mostly generated by the plasma current itself through a process called the 'internal dynamo effect", and reverses direction near the outside of the plasma (see Figure 5-1). Also, the toroidal and poloidal magnetic field have approximately the same strength within the plasma (the toroidal field is much larger in a tokamak). As a result, the total magnetic field is helical with the twist (q, number of toroidal turns per poloidal turn), much smaller than that of a tokamak. Furthermore, the current limit of tokamaks does not apply (at the cost of adding a close conducting shell) so larger plasma current densities can be supported for a given toroidal field. Most RFP machines have a moderate aspect ratio of A~4-6, with a theoretical 3 limit of -50%.

The advantages of an RFP for a fusion reactor are the following [5.1,5.2,5.3]: 1) the high current densities allow ohmic heating to ignition without complex auxiliary heating systems; 2) the relatively high beta and self-generated internal magnetic field means that the applied fields can be low and therefore magnets need not be superconducting; and 3) the high beta allows a high power density (10-20 times that of a tokamak) so that the fusion reactor core can be made smaller, less expensive and faster construction and replacement. With RFF's, unlike tokamaks, it appears possible to radiate a large fraction of the plasma power uniformly onto the first wall; high power density operation, therefore, should be possible without exceeding divertor material limits. The major disadvantages of RFP reactors is their lower energy confinement compared to tokamaks of comparable power density, and the apparent need for a conducting shell close to the plasma.

A variant of the RFP is the OHTE, which uses additional helical toroidal coil windings to form or control the toroidal field reversal. In RFPs, this toroidal field reversal is believed sustained by plasma turbulence. Therefore, the OHTE could have lower plasma turbulence and consequently improved confinement (experiments have been ambiguous). The added helical field makes access more difficult than in RFPs and requires greater precision in manufacturing, as with stellarators compared to tokamaks.

Table 5-1 lists the characteristics of the major RFP experiments and concepts. Figure 5-2 illustrates the size of the existing and next-step machines relative to reactor concepts.

TECHNICAL DESCRIPTION

Stability/Beta RFPs can operate at relatively high beta compared to tokamaks; present experiments have poloidal beta ~10-20% (the total beta is about half this), which is about that needed for reactor designs. Maintaining this beta at the higher plasma currents required for reactors is a key 25. issue. The RFP plasma exists in a near-minimum energy state, which is stable to ideal MHD and resistive tearing modes if surrounded by a conducting shell. The plasma may be unstable with a time constant which is approximately that of the shell [5.1]. Until recently, all RFPs were surrounded by shells with time constants longer than that of the discharge [5.2]. This is not possible for reactors, and is a key issue. In particular, how resistive a shell can one have and still maintain adequate stability, what is the best balance between the shell time constant and the external control field penetration time, and can instabilities be actively detected and controlled?

Confinement/Transport The mechanism that governs transport within an RFP is not well understood, nor is the mechanism that leads to the main toroidal field formation and reversal (the dynamo effect). The latter may be due to tearing modes within the plasma, which result in some level of plasma turbulence and random fluctuations of the field, and possibly a high-energy electron tail [5.15]. The fluctuations, which affect transport, are anticipated to decrease with hotter plasmas [5.15]. A major disadvantage of RFPs with respect to tokamaks is the higher loop voltage. For an ohmically heated, high power density (and therefore small) plasma, this implies a small energy confinment parameter. RFP energy confinement seems to increase 1 15 2 with plasma current, following tE(s) ~ 0.05 l0 " (MA) a (m) / Zef) [5.3,5.17].

Heating/Current Drive With no known current limits, unlike tokamaks, RFPs can be ohmically heated to ignition [5.2]. But since a toroidal current is needed for the poloidal field, some means of current drive is needed for steady-state operation. With "Oscillating Field Current Drive" or "F-0 pumping", the poloidal and toroidal fields are modulated at audio frequencies and small amplitudes (< 1 %). Some experiments have been done on ZT-40M, but the results were inconclusive [5.2, 5.4]. A lower resistance plasma (larger, hotter) is needed for more definitive tests. Alternately, radiofrequency approaches similar to tokamaks might be applied, but with some difficulty due to the relatively low field and high density [2]. The Oscillating Field Current Drive approach is preferred as it is simpler and less expensive [5.2,5.3]. Its projected efficiency is 0.3 A/W [5.3].

Impurity Control/Fuelling Present experiments use Inconel liners, the next generation of machines under construction (ZTH and RFX) use carbon tiles, and reactor designs use high Z materials [5.7]. RFPs do not have to operate at high power densities, but do so by economic and operational choice. Unlike tokamaks, it is suggested that RFPs can operate with a radiative core without significant loss of energy confinement, as has been shown in initial tests on ZT-40M [5.4]. Therefore, doping the plasma with trace amounts of high-Z elements can result in uniform radiation of the plasma power onto the first wall (e.g., only 5% of plasma power goes to the divertor in TITAN [5.3]). This highly-radiating plasma is critical for high-power-density RFP reactors, but even if not achievable, RFP's can operate at lower power densities with many of the RFP advantages intact (e.g., low coil fields, ohmic heating, simpler current drive). Reactor studies have considered both pump limiters at lower power densities (CFRPR [5.5]) and toroidal field divertors at high power densities (TITAN [5.6]). A toroidal rail limiter was installed in ZT-40M, but the limiter produced asymmetric wall currents that prevented the RFP 26. from even being formed [5.6]. Fuelling should generally be comparable to tokamaks. The RFP plasma is at a higher density, but has a smaller radius and lower magnetic fields. Pellet injection should therefore be comparable to tokamaks, except that the turbulent RFP core should provide better mixing if the pellet only penetrates a short distance. Experiments have been conducted on ZT-40M, resulting in a density increase [5.6].

Magnet/Power Supplies The high beta, self-generated plasma fields (the external toroidal field is increased five times in the plasma), and large aspect ratio all lower the magnet coil field relative to tokamaks. In particular, non-superconducting coils can be used - either copper coils or possibly even liquid metal from the blanket [5.3]. However, RFPs are sensitive to magnetic field errors, such as from perturbations due to ports and even from magnetic materials outside the machine. Therefore while the coils themselves can be simpler and less-stressed than in tokamaks, greater care in overall machine design is needed to minimize field perturbations. Power supply requirements are not obviously different from tokamaks, even though resistive magnets are used.

Reactor Engineering The nuclear design is dominated by the higher power densities, up to 20 MW/m2 have been considered [5.3] compared with 3-4 MW/m2 typical for conventional tokamaks (perhaps 8-10 MW/m2 for second-stability tokamaks) and comparable to that in fission reactors. At high- power densities, design choices are more limited - solid breeder blankets are not possible, for example. The weak edge magnetic field favors the use of liquid metals, even in high-heat flux components. For given material fluence limits, the higher power density requires more frequent blanket replacement; the possibility of single-piece maintenance of the fusion core should minimize the downtime and allow off-site assembly and easier accommodation of design improvements. The RFP has several safety advantages compared to tokamaks. In particular, the RFP density limit does not seem to cause major disruptions, as it does in tokamaks. Furthermore, the stored toroidal field energy is much smaller than in tokamaks, and the problem of superconducting coil quenching can be completely avoided by using resistive coils. The balance of plant is generally similar to tokamaks, with a large power output and standard steam-electric power systems. However, the fusion core may be a smaller fraction of the overall cost in a high-power-density RFP than in a conventional tokamak, which would result in a shorter construction time and lower overall cost.

PROGRAM PLANS

Recent progress has been slowed by budget cuts, with the new large machines (RFX, Italy and ZTH, US) delayed, OHTE (US) and HBTX (UK) shut down, and ZT-40M (US) understaffed. However, MST (US) is now running, RFX should startup in 1991 and ZTH is expected to start in 1993.

The key issues are the effect of thin shells on RFP stability, and confinement scaling to higher currents and larger machines. The new larger machines (MST, RFX and ZTH) should address these issues. Of practical concern to these new machines are the control of field errors and 27. remote tile replacement (due to limited port access). Smaller parallel machines could address more fundamental questions such as understanding the beta limit and the internal dynamo, reducing the loop voltage, and current drive [5.2].

Beyond these, one can consider either a longer-pulse/diverted machine, or possibly even a tritium-burning experiment. Such a next-step after ZTH, called ZTI, has been sketched out in a very preliminary way [5.8, 5.18].

REFERENCES

5.1 H. Dreicer, 'A Short Introduction to the Status and Motivation for and Compact Toroid Research', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V1, pg 31, 1987.

5.2 H.A.B. Bodin, 'Introduction to Reversed Field Pinch1, in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V1, pg 3, 1987.

5.3 F. Najmabadi, R.W. Conn, R.A. Krakowski et al, 'Engineering and Physics of High- Power-Density, Compact, Reversed-Field Pinch Reactors: The TITAN Study', Fusion Technology 1988, A.M. Van Ingen et al (eds.), Elsevier Science Publishers B.V. (1989) p.1779 (Proc. 15th SOFT, Utrecht, 1988).

5.4 K.F. Schoenberg et. al., 'Oscillating Field Current Drive Experiments in a Reversed Field Pinch', Phys. Fluids, 31(8), 2285, 1988.

5.5 R.A. Krakowski, R.L Hagenson, N.M. Schnurr et al, 'Compact Reversed-Field Pinch Reactors (CRFPR)', Nucl. Eng. and Design/Fusion 4 (1986) 75.

5.6 G.A. Wurden et.al., 'RFP Experiments: Results from ZT-40M and ZT-P', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V1, 159, 1987.

5.7 P.I.H. Cooke et. al., 'Engineering Design of the TITAN-II Diverter', Fusion Eng. and Design 9, 417, 1989.

5.8 C.G. Bathke, R.A. Krakowski, R.L. Miller and K.A. Werley, "The Reversed-Field-Pinch (RFP) Fusion Neutron Source: A Conceptual Design', J. Fus. Eng., V8(3) (1989) 249.

5.9 H. Dreicer, 'The Rational for the Design of CPRF/ZTH, the next step in the US RFP Program', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V1, 359,1987.

5.10 H.A.B. Bodin, 'Results from the HBTX Experiment', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V1, pg 307, 1987. 28. 5.11 C.G. Bathke et. al., 'Conceptual Design of a Reversed-Field Pinch Fusion Neutron Source', Fusion Eng. and Design V10, 39, 1989.

5.12 G. Malesani, 'The RFX Experiment', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V1, 331,1987.

5.13 S. Ortolani and M. Valisa, 'Results of Recent Studies on Eta-Beta IP, in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V1, 283, 1987.

5.14 K.I. Sato et al, 'Experimental Studies of STP-3(M) reversed field pinch in high current density regime', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proceedings of a Workshop at Varenna, Italy, V2,1063,1987.

5.15 J. Sheffield, 'Summaries of the 12th IAEA Inter. Conf. on Plasma Pysics and Controlled Nuclear Fusion, Nice, France, 12-19 October 1988', Nucl. Fusion 29(3) (1989) 499.

5.16 'World Survey of Activities in Controlled Fusion Research, 1986 Edition', IAEA, Vienna (1986).

5.17 J.W. Conner and J.B. Taylor, Phys Fluids 27 (1984) 2676.

5.18 C.G. Bathke, R.A. Krakowski, R.L Miller and K.A. Werley, 'ZTI: Preliminary Characterization of an Ignition Class Reversed-Field Pinch', Proc, 9th ANS Top. Mtg. on Tech. of Fusion Energy, Oak Brook, IL (1990 October 7-11). 29.

Table 5-1. Characteristics of major existing and proposed RFP machines*

Facility/Location Year R a I U. xE T((0) P«i« Comments/Status (m) (m) (T) (MA) (s) (1/m3) (s/m3) (keV) (MW)

Present/completed ZT-40M, Los Alamos 1984 1.14 0.2 0.2 0.34 0.004 20 8e19 2e16 0.35 0 TPE-1RM, Japan 1986 0.72 0.135 0.3 0.13 0.001 10 3e19 1e16 0.4 - OHTE, General Atomic 1986 1.24 0.19 0.24 0.4 0.005 15 1e20 2e16 0.1 - [5.16] ETA-BETA II, Padua 1987 0.65 0.125 0.3 0.15 0.001 10 1e20 1e16 0.3 0 [5.13] HBTX-1B, Culham 1987 0.8 0.26 0.15 0.22 0.01 15 4e19 5e15 0.4 0 [5.10] ZT-40M, Los Alamos 1987 1.14 0.2 0.2 0.2 0.005 - 1.5e19 2e16 1 0 Hot-ion mode, 1.5e8 n/shot [5.6] ZT-P, Los Alamos 1987 0.45 0.07 - 0.09 0.001 - 1.5e20 2e16 0.25 0 Air-core transformer [5.6] STP-3M, Nagoya 1988 0.5 0.083 0.04 0.10 0.001 - 5e19 4e16 0.3 - [5.14] REPUTE-1, Tokyo 1988 0.82 0.27 - 0.3 0.001 - 3e19 - 0.3 - MST, Wisconsin - 1.4 0.3 0.08 0.35 0.03 - 2e19 4e15 0.3 - 1989 startup [5.16] RFX, Padua 2.0 0.5 0.7 2.0 0.25 10 1e20 3e17 1 0 1991 startup [5.12] ZTH, Los Alamos - 2.4 0.4 1 1.7 0.25 10 1e20 2e17 1 0 1993 startup [5.9]

Next-step OHTE-I, General Atomic - - 0.3 - 7 1 - - - - - Ignited, 1980 concept ZTI, Los Alamos - 2.4 0.35 - 9 - 15 5e20 1e20 5 0 Ignited, 1988 concept [5.11] FTF, Los Alamos - 1.8 0.3 4.5 10.2 ss 10 7e20 3e19 9 0 Neutron source, 1989 concept [5.8]

' Simultaneous values (some inferred). Specific dates are given for achieved values; no date is given for machines proposed or under construction. is average toroidal field within plasma. tMM is flattop time of discharge, not total discharge time. Toroidal Held windings Poloidal field windings

Magnetic field lines

SOURCE: Adapted from National Research Council, Physics Through the 1990s: Plasmas and Fluids (Washington, DC; National Academy Press. 1986).

Be

Figure 5-1 Schematic diagram of a Reversed-Field Pinch illustrating the change in magnetic field direction with radius, where B(J» and B0 are the toroidal and poloidal fields, respectively [5.1]. MAJOR RADIUS (m)

Figure 5-2. Relative size and plasma current for existing and proposed RFPs. Solid shapes are actual experiments. 32.

6. DENSE Z-PINCH

INTRODUCTION

The Z-pinch is one of the oldest devices used in plasma physics research. The basic concept consists of a cylindrical column of hydrogen subjected to a current surge down the column axis. The current heats the plasma, and forms a poloidal confining magnetic field. Figures 6-1 and 6-2 illustrates the concept.

The original z-pinch started with a gas column at low pressure. The plasma initially formed near the container walls and then imploded under the JxB forces, sweeping up and ionizing the initial gas fill. The duration of the compressed state was limited by rapid MHD instabilities. The gas-embedded pinch tried to avoid this imploding phase by starting at high-pressure with an initial central plasma channel formed by a laser or electron beam. However, the continual accumulation of additional background gas into the plasma prevented achieving significant conditions. In the high-density Z-pinch, the gas is replaced by a solid fibre and surrounded by a vacuum. Under these conditions, the discharge has proven to be unexpectedly stable up to high ion densities and significant temperatures. The discharge is characterized by a limiting current at which the ohmic heating and radiation losses are balanced. At higher currents, radiation would exceed heating and the pinch would collapse. This "Pease-Braginskii" current is ~1.5 MA in a hydrogen pinch, higher in a DT pinch with the additional alpha-particle heating.

The main advantages of a Z-pinch are its simple geometry, lack of magnetic coils, and small size. The main disadvantages are the possible instability under reactor conditions, and the technology to supply high-rates of short-pulse electrical power to the pinch.

Table 6-1 summarizes the parameters associated with completed and planned experiments.

TECHNICAL DESCRIPTION

Stability/Beta The key question for the Dense Z-pinch is its stability. Classical Z-pinches with imploding current sheaths were highly unstable. Ideal MHD theory predicts stability to the m=0 sausage instability if the current profiles are sufficiently diffuse. Relatively recent experiments with solid-density pinches and different formation methods have shown unexpected stability; all confirm the absence of kink (m=1) instabilities, and the onset of the sausage (m=0) instability appears to be delayed by ~ 100 ns at currents approaching 1/3 of the Pease-Braginskii limit. The absence of the m=0 instability can be explained by pressure or current profile effects, but the absence of the m=1 instability is surprising. It is not clear if the instability is delayed until dl/dt=O (as suggested by NRL), and/or is due to resistive, viscous, skin or axial-flow effects [6.7]. In present experiments, the observed stable period is about 100 times the Alfven (MHD instability) time; stability for ~4000 times the Alfven time are needed for practical neutron source [6.7]. 33.

Confinement/Transport The pinch is a short-pulse plasma surrounded by vacuum. The major energy loss appear to be radiation, with some axial conduction. The loss rate is such that the energy confinement time appears to be longer than the stable discharge period.

Heating/Current Drive The plasma is ohmically heated to ignition by the central current which is directly obtained from the power supplies. The current transfer interface between the plasma chamber and the insulated power supplies is a sensitive engineering region. Present water/vacuum interfaces work reasonably well, but reactors will require innovative solutions such as use of liquid metal jets as the electrodes in order to position the actual interface far away from the DT reactions.

Impurity Control/Fuelling Impurity control is not the issue as the discharge is over before there is significant interaction with the structure. Fuelling is presently accomplished by extruding cryogenic fibers of frozen D2. The cryogenic apparatus is simple and reliable, producing fibers of about 50 urn in diameter and several cm in length. Fiber formation will be more difficult for reactors, where they must be fed at a high rate and their source must be shielded from the fusion reaction. The use of fiber 'injectors' at 40 m/s has been suggested but not analyzed in any detail. A chamber vacuum of < 0.01 Pa is needed to insulate the electrodes, rather than for impurity control. The associated vacuum pumping (e.g., ~60 m3 He/s at 300 MWf [6.4]) may constrain the allowable pulse repetition rate. There is no plasma in long-term contact with the walls, so the plasma facing components are not critical to present low-repetition rate experiments. The need to directly conduct current into the plasma, and the short-pulse nature of the pinch suggests that liquid metal coolants and inertial-fusion liquid-metal first walls, would be most suitable for reactors.

Magnet/Power Supplies The Z-pinch does not require an external magnetic field. The pulsed power supplies for present experiments are fairly conventional: oil-dielectric capacitor banks discharging into pulse-compressing water-dielectric capacitors, which in turn discharge into the fiber. The main issues will be developing 10-100 Hz repetition rate power supplies, and in hardening the interface between the power supply and the pinch chamber. Much of the energy goes into forming very large (1 -10 kT) magnetic fields which constrain the pinch, rather than into plasma kinetic energy. This magnetic energy must be recovered back into the capacitor banks at the end of the pinch pulse. In present experiments, it is simply routed into clumping resistors.

Reactor Engineering Dense Z-pinch machines are relatively compact and moderate power output devices with limited stored energy per device. Most of the initial input energy is stored as inductive magnetic energy (e.g. 7.2 out of 7.6 MJ for a 300 MWf reactor), and must be efficiently recovered. Fortunately, it can be recovered directly back throught the power supply circuit with high efficiency (In other fusion concepts, the recirculating power must usually be recovered thermally). The fusion power itself would be recovered through inertial-fusion liquid- 34. metal blankets, using conventional thermal power cycles. The Dense Z-Pinch appears limited to moderate power outputs, perhaps 100 MWe. This makes them of interest for such applications as naval propulsion or perhaps space missions. Large-scale electricity production would be accomplished by a series of such units. The ability of dense Z-pinches to ignite advanced fuels is not clear.

PROGRAM PLANS

The recent and remarkable absence of plasma instabilities when using solid deuterium fibres at cryogenic temperature has led to a renewal of interest in Z-pinches. These results were obtained in three machines operating at about 20-40% of the Pease-Braginskii current. The key issue for the Dense Z-pinch is to determine how thw stability scales at the Pease- Braginskii current limit. This is critical to the feasibility of the Dense Z-Pinch and must be experimentally demonstrated as it is not theoretically understood. This issue is being addressed by the current series of experiments.

These experiments are all designed to operate near the current limit, but differ in the current ramp rate: ZFX at NRL ~4 kA/ns to 1.6 MA ZEBRA (HDZP-II) at LANL ~6 kA/ns to 1.2 MA MAGPIE at Imperial College -12 kA/ns to 1-1.5 MA HDZP at KfK KALIF facility The NRL group assumes that the same dl/dt as their successful first experiment is best for stability; the LANL group assumes the same stability duration as their first experiment, and therefore use a higher dl/dt. The NRL and LANL machines are just starting operation, and definitive information on stability scaling may be available by the end of 1990. The Imperial College facility is intended to study the physics of radiative collapse when the current limit is exceeded. First experimental results are expected in early 1992.

Assuming that the pinches remain reasonably stable and useful temperatures are achieved, the next step would be to conduct DT tests. The actual amount of tritium needed in these first tests is quite small (~10-100 Ci), and may be conducted with the existing machines.

REFERENCES

6.1 A.H. Glasser, 'Z-Pinch and Plasma Focus*, Nuclear Fusion 29 (1989) 129.

6.2 P.R. Parker et al, 'Plasma Physics and Controlled Nuclear Fusion Research', Nuclear Fusion 29 (1989) 489.

6.3 A,E. Robson, The Dense Z-Pinch', Nuclear Fusion 29 (1989) 1825.

6.4 A.E. Robson, The Dense Z-Pinch in a Small Fusion Reactor', Proc. Inter. Conf. Dense Z-Pinches (AIP Conference ... 195 (1989)). 35. 6.5 D.H. Crandall, et al, 'Reversed Field Pinch, Compact Toroid, and Dense Z-Pinch", J. Fusion Energy 8 (1989) 9.

6.6 A.H. Glasser et al, 'The Physics of the High Density Z-Pinch, Proc. 12th IAEA Inter. Conf. on Plasma Pysics and Controlled Nuclear Fusion, Nice, France, 12-19 October 1988, IAEA, Vienna (1989), Vol.2, 557.

6.7 R.A. Krakowski, J.D. Sethian and R.L. Hagenson, The High-Density Z-Pinch as a Pulsed Fusion Neutron Source for Fusion and Materials Testing', Jrnl Fusion Energy 8(3/4) (1989) 269.

6.8 M.G. Haines et al, 'Z-Pinch Equilibria and Stability: Experiment and Theory', Proc. 12th IAEA Inter. Conf. on Plasma Pysics and Controlled Nuclear Fusion, Nice, France, 12-19 October 1988, IAEA, Vienna (1989), Vol.2, 565. 36.

Table 6-1. Data on existing and planned dense Z-pinch devices*

Parameter NRL LANL KfK Imp. Coll.

Machine ZFX HDZP HDZP- FTF MAGPIE Date 1988 - 1988 - 1988 1988 - Fiber compositicri D2 CD2 Fiber form solid solid solid solid solid solid gas solid Fiber diameter, urn 80 80 20-40 30 30 <20 20000 20 Fiber length, cm 5 5 5 5 10 3-5 5 5 Capacitor energy, kJ 30 350 40 200 540 220 300 Pinch voltage, MV 0.5 0.75 0.6 3.2 10 1.5 2.4 Peak current, MA 0.64 1.6 0.25 1.2 1.9 1 0.15 1.5-2 Stable discharge, ns 150 630 70 100 100 140 60 150 Ion temp., keV 0.2 0.2 10 15 0.15 - Ion density, 1/m3 1e27 2e27 1e28 3e28 1e28 1e25 - /pulse 1e8 1e16 2e17 2e9 nx, s/m3 1e19 1ei9 1e20 3e20 4e18 - , kT 0.6 0.6 6 10 Reference [6.7] [6.6] [6.6,6.7] [6.7] [6.8] - ' Simultaneous values (some inferred). No dates are given for machines under commissioning or proposed. Magnetic field is estimated based on peak current and some radial expansion, average field across pinch is estimated as half the peak (outer) value. m is estimated assuming peak (power producing) conditions last for 1/10 of discharge time. INDUCED MAGNETIC FIELD PINCH CURRENT

ELECTRODE ELECTRODE

CAPACrrOR POWER SUPPLY

Figure 6-1. General schematic of Z-pinch

Figure 6-2. Illustration of LANL second generation HDZP [6.2] 38.

7. FIELD-REVERSED CONFIGURATION

INTRODUCTION

A Field-Reversed Configuration (FRC) is a plasma torus, confined by a poloidal magnetic field contained within an external axial field, as shown in Figure 7-1 [7.3]. The poloidal field of the torus reverses the direction of the external field along the axis of the external field. The FRC magnetic configuration is similar to the Spheromak, except that the latter also contains a toroidal field within the plasma torus. Field-Reversed Mirrors can have toroidal fields anywhere from none, as in FRC's, to large, as in Spheromaks.' All are generically referred to as Compact Toroids. Physically, the FRC torus is elongated along the machine axis ("prolate"), while the Spheromak is roughly circular ("oblate").

An FRC can be generated by fast reversal of the external current of a theta pinch during its formation phase (Field-Reversed Theta Pinch), or by creating an internal current ring around the axis of a mirror field, which forms a poloidal field reversing the main mirror field (Field- Reversed Mirror). Other methods include coaxial guns and rotating magnetic fields. Most present FRC's are of the theta-pinch type [7.3].

The primary advantages of the FRC are its very high beta, lack of internal coils, and natural divertor. The high beta (~90%) allows for high power density (and hence small sizes and the Compact Toroid name), lower field magnets and possibl" advanced fuels. (The Spheromak does not share this high beta capability.) The lack of internal coils allows small plasmas (e.g. no space needed for magnets and shielding), simpler geometry (e.g , no interlocked vacuum vessel and coils), and a movable plasma [7.9]. The external fields provide a natural divertor to isolate the torus from the walls, removing reaction products, and possibly provide for direct conversion of the fusion power.

The FRC requires an internal toroidal current to maintain its poloidal confining field. This current is produced during formation, and its subsequent resistive decay limits the lifetime of the torus. The energy confinement is of the same order as this magnetic-field decay time. Most reactor studies consider pulsed, translating toroids, but steady-state beam-sustained FRCs have also been proposed.

Table 7-1 lists the parameters of most FRC experiments. Figure 7-2 illustrates the relative size of the present machines and of reactor concepts.

TECHNICAL DESCRIPTION

Stability/Beta 2 FRC's have a very high equilibrium beta, ~ 1 - {rjrc) l2, where rs is the separatrix radius and rc is the conducting (flux-conserving) wall radius. Values of -90% are readily achieved in experiments. However, the plasma is theoretically prone to various instabilities. For many years, FRCs suffered from an n=2 rotational instability that distorted the plasma cross-section 39. and caused it to touch the wall within tens of microseconds. The n=1 tilt mode was similarly expected to be fatal, but had not been seen experimentally. The use of weak external coils, careful formation and an elongated plasma have controlled these instabilities in present experiments, whose plasmas last for times consistent with resistive decay of the fields [7.3]. However, it is generally believed that the absence of the tilt instability is a consequence of the small size of the experiments to date, where the plasma minor radius is not much larger than the average plasma ion gyroradius. This ratio, s, is typically less than 2, thus introducing a helpful averaging or 'finite Larmor radius' effect. For fusion reactors, s ~ 30 is expected and the tilt instability (theoretically expected for s > 3) could destroy the plasma. This is the critical feasibility issue for FRC's.

Recent experiments on FRX-C/LSM may have seen indications of this instability at s~2 [7.10]. The newest machine, LSX, is designed to achieve s ~ 8 and so address this issue. It has been shown theoretically that adding a circulating ion ring will stabilize the plasma, and also suggested that the hot fusion reaction products in a fusing FRC will have small values of s and therefore also stabilize the plasma. Also, plasma elongation, profile and edge effects, conducting boundaries, and plasma rotation are stabilizing factors that may contribute even in large plasmas [7.3].

Confinement/Transport Experiments so far are physically small with overall lifetimes comparable to the particle and energy lifetimes. Energy confinement is anomalous in present experiments, but appears to be dominated by anomalous particle transport due to large ion orbits in the low-field plasma center and to large density gradients at edge. Ion thermal conduction appears to be small because of the large region between the separatrix and first wall in present experiments. 2 5 Present experiments seem to follow a tE(s) ~ 0.01 R(m) Bexlernai(T) / Tj(keV) °- scaling [7.3]. It has been suggested that about a factor of ten improvement would be desirable for a reactor [7.3]. The increased plasma size, if stable, should result in improved confinement and less edge effects. The scaling of confinement to larger plasmas (larger s) is a key issue.

Heating/Current Drive Experimental FRCs have been pulsed, and heated by shocks, ohmic power and compression. Therefore, reactor designs tended to assume pulsed FRCs, heated to ignition by compression (e.g., CTOR). However, more recent assessments have considered the use of particle beams for stability, heating and steady-state operation [7.3]. First beam experiments (in 1979) in a field-reversed mirror configuration on the 2XIIB machine were unsuccessful. Neutral beam injection experiments have recently started on FIX [7.11].

Impurity Control/Fuelling For the physically small FRC plasmas produced to date, impurities are significant, although relatively hot and clean plasmas have been formed. Impurity control is important, and should be achievable through careful formation techniques, improved materials, and the natural FRC divertor. FRCs have a high plasma power density, which will constrain the plasma-facing component design for longer-lifetime plasma, but moving the plasma should lower the average heat loads. Present short-duration experiments have been fuelled by the initial particle load. Fuelling by moving the torus over a pellet will be attempted in 1990 on FRX-C/Compressor. 40.

Magnet/Power Supplies Reducing the cost of reactor power supply and of the initial impurities depends critically on developing slow formation techniques that will allow the use of rotating generators rather than high-voltage capacitor banks. This is a key issue. This will probably be more difficult than with spheromaks. The magnets are much easier for FRC's than tokamaks due to the high plasma beta which allows lower magnetic fields at the conductor (and therefore lower structural forces and greater material choice), and due to their simple solenoidal geometry.

Reactor Engineering The nuclear design is simpler relative to tokamaks since the chamber is cylindrical with the blanket and shielding entirely outside the plasma torus, and since the plasma torus can be moved which allows the formation region to be separate from the burn region. It is more difficult in that it is a high power density system. The high power density and small aspect ratio may allow smaller fusion power outputs compared to tokamaks. The high beta and ease of direct conversion also makes this concept very attractive for advanced fuels.

PROGRAM PLANS

Experimental work on FRC's first began in the 1960's within the USSR and USA. Subsequent research was also undertaken in the UK and FRG; these latter projects have since been discontinued but Japan has entered with a modest program within the university sector. The US has the largest machines. The US and Japan host a joint annual meeting on FRCs.

In recent experiments at FRX-C (1989), the FRC was compressed to higher fields with a consequent increase in temperature, resulting in one of the hottest FRC's. Other experiments on FRX-C have provided first indications of the tilt instability at s - 2. The LSX experiment is just starting operation, and can test stability and confinement for plasmas up to s ~ 8. Determining the existence and, if necessary, control of the tilt instability is the critical question for FRC's.

Next-step experiments should be physically large enough to produce relevant plasmas. As with LSX, the primary topics of interest are stability and transport scaling. A particular suggestion would be for an FRC with particle beams to study stability, and possibly heating and/or current drive (full heating would take considerable beam power). Such an experiment could be similar in scale to LSX. A new FRC experiment is under consideration in Japan. There are no plans for a near-term tritium experiment.

REFERENCES

7.1 S. Ortolani and E. Sindoni, Physics of Mirrors, Reversed Field Pinches and Compact Tori, Societa Italiana de Fisica, Bologna, Italy (1988).

7.2 A.L. Hoffman, 'Formation of Field-Reversed Configuration Using Scalable Low-Voltage 41. Technology' Fusion Technology 9, 48 (1980).

7.3 M. Tuszewski, "Field Reversed Configurations', Nuclear Fusion 28(11), 2033 (1988)

7.4 L. Steinhauer, Spectra Technologies (Bellevue, Washington), private communication, February 1990.

7.5 A.C. Kolb et al, 'Plasma Confinement, Heating and Losses in Pharos with an Extended Current Pulse1, Proc. 2nd Inter. Conf. on Plasma Physics and Cont. Nucl. Fusion, Culham, 1965, Vol.1 IAEA Vienna (1966) 261.

7.6 'World Survey of Activities in Controlled Fusion Research, 1986 Edition', IAEA Vienna (1986).

7.7 J. Sheffield, 'Summaries of the 12th IAEA Inter. Conf. on Plasma Pysics and Controlled Nuclear Fusion, Nice, France, 12-19 October 1988', Nucl. Fusion 29(3) (1989) 499.

7.8 G. Vlases, Proc. 5th Symp. on Physics and Tech. of Compact Toroids, Bellevue, Washington, 1983 January.

7.9 D. Rej et al, 'Experimental Studies of field-reversed configuration translation', Phys. Fluids 29(3) (1986) 852.

7.10 D.J. Rej, G.A. Barnes, M.H. Baron et al, "Flux Confinement Measurements in Large Field-Reversed Configuration Equilibria', Phys. Fluids B (1990).

7.11 S. Goto et al, 'Large Volume FRC Plasma Production and High Energy Particle Injection Experiments in FIX Machine', p.107 in Proc. US-Japan Workshop on Field- Reversed Configurations with Steady-State High-Temperature Fusion Plasmas and the 11th US-Japan Workshop on Compact Toroids, LA-11808-C, Los Alamos, 1989 November 7-9.

7.12 D.J. Rej et al, 'Initial Results from FRC Compression Experiments on FRX-C/LSM', p.98 in Proc. US-Japan Workshop on Field-Reversed Configurations with Steady-State High- Temperature Fusion Plasmas and the 11th US-Japan Workshop on Compact Toroids, LA-11808-C, Los Alamos, 1989 November 7-9. 42.

Table 7-1. Major field-reversed configuration experiments*

Facility/Location Year L* B.,, 1 <8P> xE T(0) P., Comments/Status 3 3 (m) (m) (T) (MA) (s) (%) (1/m ) (s/m ) (keV) (MW)

PHAROS. NRL 1966 <1.8 0.04 4 2e-5 80 1G22 1e17 0.9 FRTP [7.3,7.5] BN, Kurchatov 1975 <0.9 <0.1 0.4 • 5e-5 - - - - - [7.3] TOR, Kurchatov 1979 <1.5 <0.15 1 - 1e-4 - - - - - [7.3] FRX-A, Los Alamos 1980 0.36 0.05 0.7 - 3e-5 • 4e21 2e16 0.2 - [7.3] TRX-1, Math Sciences 1983 0.06 1 - 85 3e21 1e17 0.25 - [7.8] FRX-C, Los Alamos 1983 <2 0.11 0.8 1.3 2e-4 92 2e21 8e16 0.6 2.5 [7.3,7.6] TRX-2, Spectra Tech 1986 <1 0.06 1.3 1 1e-4 88 5e21 2e17 0.6 - FRTP [7.3,7.6] ROTAMAK, Australia 1986 <0.5 0.1 - - 0.02 - - - - 0.003 [7.3,7.6] OCT-S, Osaka 1986 0.6 0.05 0.2 0.6 1e-4 80 2e21 1e16 0.1 - FRTP [7.3,7.6] PIACE, Osaka 1986 <1 0.04 1 0.5 6e-5 90 3e21 3e16 0.4 - FRTP [7.3,7.6] NUCTE. Nihon 1986 <2 0.03 1 - 5e-5 - 3e21 4e16 0.2 - FRTP [7.3,7.6] CTTX, Pennsylvania 1986 <0.5 0.03 - - 5e-5 95 2e21 4e16 0.03 - FRTP [7.6] FRX-C/LSM, Los Alamos 1988 1.8 0.18 0.4 0.8 2e-4 87 6e20 4e16 0.4 - FRX-C/Comp, Los Alamos 1989 0.8 0.06 1.8 1 - - 3e21 1e17 1.4 600 108n/pulse[7.12] FIX, Osaka 1989 3 0.2 0.05 • 4e-4 - 5e19 - 0.07 - [7.11] LSX, Spectra Tech 2 0.3 0.75 - 0.001 - 2e21 2e18 0.3 - [7.3] FRX-D, Los Alamos 0.7 0.18 1.4 6e-5 90 5e21 - 0.5

* Simultaneous values (some inferred). Specific dates are given for achieved values; no date is given for machines proposed or under construction. L is the plasma length, most reports only give the coil length; r, is the separatrix radius;

B,x, is the external vacuum magnetic field; is tne fattop time. SEPARATRIPART X THCTA-PINCTHCTAP H COIL

/^/////////////////^^

/ C Z> N :> J

\ CLOSED POLOIOAL OOPEN MAGNETIC DISCHARGE TUBE MAGNETIC FIELD LINE FIELD LINE

FIELD-REVERSED THETA PINCH J\ |R B9

Figure 7-1. Schematic diagram of a Field-Reversed Configuration (FRC) and its magnetic field configuration, where Be is the poioidai field [7.3]. 8

TRACT P 130 MWe

LU LL o CTOR LU z 310 MWe (3 4 PHAROS

< DC LU F.RX»C/Comp LU FRX-D LSX

•FRX-C/LSM OCT c > FIX 0 2.5 50 7.5 10 PLASMA LENGTH (m)

Figure 7-2. Relative size of present and projected reactor FRC plasmas. (No ignition steps are shown.) Solid lines are actual experiments. 45. 8. SPHEROMAK

INTRODUCTION

A Spheromak is a plasma torus confined by poloidal and toroidal magnetic fields contained within an external axial field, as shown in Figure 8-1. The poloidal field of the torus reverses the direction of the external field along the axis of the torus. The Spheromak magnetic fields are similar to the FRC, except that the latter does not have a toroidal field within the plasma torus. Both are generically referred to as Compact Toroids. Physically, the FRC torus is elongated along the machine axis, while the Spheromak is roughly circular in cross-section. The poloidal and toroidal magnetic fields of the Spheromak are roughly equal.

The advantages of the Spheromak are the absence of internal coils, natural divertor, and moderate beta. The lack of internal coils allows small plasmas (e.g. no need to allow space for internal magnets and shielding), reduces engineering complexity (e.g., interlocked vacuum vessel and coils), and allows the plasma to be moved. The moderate beta (up to 20%) allows for moderate power density (and hence small sizes for given power), and lower field magnets. The field geometry has a natural divertor for removing reaction products, and possibly provide for direct conversion of the fusion power. The major concern with Spheromaks is their lifetime and confinement scaling to fusion plasmas.

A major advantage relative to FRCs is that both the shift and the tilt instability have been observed and suppressed. In the FRC, for comparison, the tilt instability has not yet been seen and so it is not clear if it can be readily stabilized. The major disadvantages relative to the FRC are the lower beta (typically <6>~10% rather than up to 90%), the production of impurities from the electrodes usually used in Spheromak (but not FRC) formation, and the possible need for a close fitting conducting shell for stability. Like RFPs, Spheromaks have no minimum-B well and are stabilized by magnetic shear (tokamaks have an average minimum-B well and moderate shear). However, the shear at the plasma boundary is much smaller in spheromaks than in RFPs.

The Spheromak, like the FRC, requires an internal toroidal current to maintain its poloidal confining field. This current is produced during formation, and its subsequent resistive decay limits the lifetime of the torus. However, it is possible that the spheromak might be refreshed (this is probably easier than with FRCs).

Table 8-1 lists the characteristics of the major Spheromak experiments. The relative size and magnetic field of present spheromaks and conceptual reactor designs is shown in Figure 8-2.

TECHNICAL DESCRIPTION

Stabilitv/Beta The spheromak equilibrium is a minimum energy state (like RFPs, but not true for FRCs). It is theoretically susceptible to tilt and shift instabilities. The n=1 tilt has been observed and suppressed by the addition of a weak external quadrupole field. The n=1 shift instability has 46. been corrected by various techniques, especially the use of flux conserving shells with or without bias fields. The need for a close-fitting shell may limit the spheromak to stationary reactor concepts. The classical spheromak has a low theoretical (Mercier criterion) plasma limit of around 1% due to the small amount of magnetic shear. If there is no current flowing along some region around the central axis of the spheromak, then higher <8> limits are possible. In experiments, the observed values of <3> are much higher, and - 20% relative to the external coil field is considered achievable, [8.2]

Confinement/Transport There is no general experimental scaling law available for spheromak energy confinement. The energy confinement in early experiments was dominated by impurity radiation, in later experiments by particle loss. In larger and more recent experiments on CTX, it has been suggested that the interaction with neutral gas at the edge, along open field lines, dominates energy losses [8.2]. In particular, the resulting low electron edge temperature could result in increased edge resistivity and therefore dissipation of the magnetic field. (In minimum energy states like the spheromak, large currents may need to flow near the edge.) The energy confinement time is related to the (fractional) plasma beta and magnetic energy resistive decay time by TE = 1.5 <0> TB2 [8.2]. Overall confinement and its scaling is the critical issue for spheromaks.

Heating/Current Drive Experimentally, ohmic heating (magnetic field decay, plasma current) has been the usual heating method. Compression (S-1, CTX) and collisional conversion of directed kinetic energy to thermal energy (RACE, TRISOPS, C0P) have also been tested. For reactors, ohmic heating to ignition may be sufficient but this is not certain, particularly with slow formation methods [8.8]. Radiofrequency heating should be suitable [8.14]. The spheromak decays with a lifetime under reactor conditions of tens of seconds. However, it seems possible to refresh the plasma current for longer operation using direct helicity injection from electrodes [8.3,8.8].

Impurity Control/Fuelling Fuelling is similar to FRCs and has not been major factors in experiments. In fact, lower densities are sometimes desired in order to achieve higher temperatures for given available power. (The formation densities are constrained by the need for sufficient initial fill gas pressure for reliable breakdown.) Pellet fuelling has been suggested (possibly by moving the spheromak onto the pellet), even for present experiments where it could be used to minimize the background gas pressure (and edge cooling) for sustained pulses. Impurity pickup during plasma formation is a major issue. The power density is higher than conventional tokamaks, unless the plasma is moved.

Magnet/Power Supplies Slow formation techniques must be used to minimize power consumption in reactors. This should be easier than with FRCs. The magnets have simple cylindrical geometry and modest field strengths.

Reactor Engineering The nuclear design is similar to FRCs with respect to geometry and pulse lifetimes. The lower 47. spheromak beta should result in lower power density than in FRCs, but higher than in tokamaks. The main engineering question is whether the spheromak is stationary or moving. If stationary, the wall loads are quite high (e.g., 17 MW/m2); if the torus is moved through a burn chamber, the heat load would be spread out (e.g., 2 MW/m2) and the plasma formation hardware would be separated from the nuclear burn region. Spheromaks have been accelerated to speeds greater than 1000 km/s. The stored magnetic energy is low compared to tokamaks. The stored energy and balance of plant would be similar to FRCs.

PROGRAM PLANS

The history of spheromak research began in the 1970's in the USA, and in the 1980's in Japan; the USSR does not seem to have a spheromak research program and the European countries engaged in only short-term activity (U. Manchester, UK and U. Heidelberg, FRG). The US DOE is still supporting small fusion spheromaks at U. Maryland and U. California (Berkeley); but the present largest and best spheromaks are at Lawrence Livermore National Laboratory (RACE), Los Alamos National Laboratory (CTX), and the US Air Force Weapons Laboratory. These latter machines are supported primarily for high velocity studies. The most recent interesting results are from the CTX project, which has achieved record temperatures although in a "poor" confinement mode.

The key issue for spheromaks seems to be generally scaling up their performance through larger experiments, possibly with emphasis on improving the beta by shaping. This concept, like the FRC, is stili several experiments removed from useful tritium fusion experiments.

At present, no group is planning to build a significant next-step fusion experiment. There is, however, interest in Spheromaks for other applications, generally involving their ability to be rapidly accelerated and compressed. In particular, small spheromaks may be quite well suited to fuelling and current drive in tokamaks (or other machines) [8.9], to inertia! fusion [8.10], and to space propulsion [8.11].

REFERENCES

8.1 G.C. Vlases, 'A Comparison of Spheromaks and FRCs', Proc. 5th Symp. Physics and Tech. of Compact Toroids, 1982 November 16-18, Mathematical Sciences Northwest Inc., January 1983.

8.2 J.C. Fernandez et al, 'Energy Confinement Studies in Spheromaks with Mesh Flux Conservers', Nucl. Fusion 28(9) (1988) 1555.

8.3 J.C. Fernandez et al, The m=1 Helicity Source Spheromak Experiment', Phys. Fluids B 1(6) (1989) 1254.

8.4 B.L. Wright et al, 'Helicity Conservation and Energy Confinement in CTX Spheromak', Proc. 11th Inter. Conf. on Plasma Physics and Contr. Nuclear Fusion, Kyoto, 13-20 48. November 1986, IAEA Vienna ^987).

8.5 G. Raupp et al, 'Investigations of Stabilized Spheromaks in Heidelberg Spheromak Experiment', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proc. of Workshop at Varenna, Italy, Vol.2 (1987) 1123.

8.6 Y. Kato et al, 'Spheromak Confinement in an Ideally-Closed Flux Conserved, in Physics of Mirrors, Reversed Field Pinches and Compact Tori, Proc. of Workshop at Varenna, Italy, Vol.2, (1987) 1137.

8.7 D.R. Wells, P.E. Ziajka and J.L Tunstall, 'Hydrodynamic Confinement of Thermonuclear Plasmas TRISOPS VIII (Plasma Liner Confinement)1, Fusion Tech. 9 (1986) 83.

8.8 R.L. Hagenson and R.A. Krakowski, 'Steady-State Spheromak Reactor Studies', Fusion Tech. 8 (1985) 1606.

8.9 L.J. Perkins, S.K. Ho and J.H. Hammer, 'Deep Penetration Fuelling of Reactor-Grade Tokamak Plasmas with Accelerated Compact Toroids', Nucl. Fusion 28(8) (1988) 1365.

8.10 M.T. Tobin, W.R. Meier and E.C. Morse, The Compact Torus Accelerator, A Driver for ICF, Fusion Tech. 10 (1986) 679.

8.11 V.E. Haloulakos and R.F. Bourque, 'Fusion Propulsion Study', AL-TR-89-005, Astronautics Laboratory, US Air Force (July 1989).

8.12 G. Vlases, Proc. 5th Symp. on Physics and Tech. of Compact Toroids, Bellevue, Washington, 1983 January.

8.13 T.R. Jarboe et al, 'Progress with Energy Confinement Time in the CTX Spheromak', p.132 in Proc. US-Japan Workshop on Field-Reversed Configurations with Steady-State High-Temperature Fusion Plasmas and the 11th US-Japan Workshop on Compact Toroids, LA-11808-C, Los Alamos, 1989 November 7-9.

8.14 E.C. Morse et al, 'Berkeley Compact Toroid Experiment: Experimental Results and Progress1, p. 148, ibid.

8.15 C.W. Hartmann et al, 'Acceleration of Spheromak Toruses, Experimental Results and Fusion Applications', p.197, ibid.

8.16 P.K. Browning et al, 'Progress on SPHEX, the Spheromak at Manchester', p.152, ibid.

8.17 M. Katsurai, N. Amemiya and A. Hayakawa, 'Experimental and theoretical studies on the magnetic configuration of bympy z-pinch/flux-core-spheromak', p. 138, ibid.

8.18 A.B. Filuk et al, 'Observations and Modelling of Electron Density on the Maryland Spheromak1, p.168, ibid. 49.

Table 8-1. Major Spheromak experiments*

Facility/Location Year R I tpulu TE Tj(O) P.,, Comments/Status < 3 (m) (m) (T) (MA) (s) (1/m3) (s/m ) (keV) (MW)

BETA-II, Livermore 1981 0.12 0.5 0.2 2e-4 2e21 FRM ProtoS-1, Princeton 1982 0.07 0.02 0.2 0.015 7e-5 10 2e20 1e15 0.03 0 [8.12] CTX, Los Alamos 1983 0.2 0.2 0.3 0.2 5e-4 7 6e19 1e15 0.09 0 [8.4,8.12] TRISOPS-VIII, U. Miami 1985 0.02 0.02 3.5 1e-5 2e22 1 Adiabatic compression [8.7] S-1, Princeton 1986 0.45 0.25 0.7 0.4 0.001 10 1e20 1e16 0.4 CTX, Los Alamos 1986 0.3 0.3 0.5 0.5 8e-4 1.8 5e19 1e15 0.08 0 [8.4] HSE, Heidelberg 1987 0.05 0.05 0.5 2e-5 20 1e21 0.02 0 [8.5] CTCC, Osaka 1988 0.25 0.2 0.1 0.09 0.001 2e19 0.03 0 [8.6] RACE, Livermore 1988 0.2 0.2 0.4 3e20 0.02 40 kJ kinetic energy [8.15] SPHEX, U. Manchester 1988 0.09 0.04 0.001 4e20 0.03 [8.16] TS-3, Tokyo 1989 0.15 0.07 1e-4 0.03 [8.17] MS, Maryland 1989 0.15 0.15 0.2 1e21 0.05 [8.18] CTX, Los Alamos 1989 0.13 0.13 0.9 3e-4 3e20 0.35 BCTX, Berkeley - 0.15 - 20 [8.14] MARAUDER, Kiitland - 0.53 0.15 0.1 HESS, Los Alamos - 0.12 0.12

* Simultaneous values (some inferred). Dates are given for achieved results. is average minor radius, (a+b)/2 is volume-average magnetic field within spheromak OPEN MAGNETIC FIELD LINES

TOROIDAL MAGNETIC FIELD AND PLASMA CURRENT

CLOSED POLOIDAL MAGNETIC FIELD LINES

SPHEROMAK

Figure 8-1. Schematic diagram of a Spheromak and its magnetic field configuration,

where B, and B9 are the toroidal and poloidal fields, respectively. TCT 500 MWf

KARIN 647 MWe

200 MWe \

0 0.5 1.0 1.5 2.0 MAJOR RADIUS (m)

Figure 8-2. Relative size of present and projected reactor Spheromak plasmas. (No ignition steps are shown.) Solid lines are actual experiments. 52.

9. IGNITION EXPERIMENTS AND REACTORS

9.1 Ignition

The parameters nx (confinement) and T (temperature) provide simple overall measures of how close a concept is to ignition, including the complete range of physics and technology needed to achieve such plasma conditions. The classical summary of physics progress is the Lawson diagram, nx versus T. Figure 9-1 shows the Lawson diagram for the experimental data reported in Sections 2 to 8 (also including some mirror and laser fusion results).

As an overall physics feasibility figure-of-merit, the product nxT has been suggested. Since breakeven requires nxT~5-10 s-keV/m3, this parameter provides a sense of how far a concept has to go. Table 9-1 summarizes the achieved nxT for the various fusion concepts. It should be noted that the values of ns x and T are not always precisely measured nor consistently reported. For exam-is, there may be significant differences between central and average values, between ion ami , and between core and global values. Nonetheless, the values are indicative of the relative standing of the various concepts.

Both Figure 9-1 and Table 9-1 indicate quite clearly that tokamaks and lasers have the best physics base. The next group consists of mirrors, stellarators and dense Z-pinches. The good performance of mirrors and stellarators is not surprising, but the inclusion of dense Z- pinches to this group reflects the recent and unexpected stability in solid-fiber based discharges. The next group consists of reversed-field pinches and compact toroids. However, although the nxT performance of RFPs and FRCs are similar, reversed-field pinches enjoy more confidence in their MHD stability and scaling to ignition.

At present, therefore, only tokamaks and lasers are reasonably established with respect to achieving breakeven and even ignition. All other concepts require improvements in performance by at least two orders of magnitude. However, there are a number of next-step machines now under construction or proposed for construction in the 1990's. If we include the predicted performance of machines now under detailed design or construction, then the performance of dense Z-pinches (HDZP-II), stellarator (LHD), RFPs (FRX, ZTH) and FRCs (LSX) jump to around 10, 0.6, 0.003 and 0.006 s-keV/1020 m3, respectively, as also shown in Table 9-1. If these projections are met, then the relative 'risk' is considerably reduced for all these candidates. The dense Z-pinch could even exceed predicted performance in the present large tokamak and laser projects, and begin tests with tritium. Only the spheromak and spherical torus do not have firm plans for the construction of machines with substantive physics performance.

In part, the better nxT performance of tokamaks and lasers reflects the much larger effort that has been spent on them. Therefore, it is interesting to compare the plasma performance with respect to machine cost. Unfortunately, unlike machine parameters, capital cost is usually loosely defined. The cost of power supplies, building and diagnostics are significant for new 'green-field' projects, but are often not included in quoted construction costs at existing facilities. Some major equipment, particular auxiliary heating systems, are purchased later 53. and again not included in quoted construction costs. Nonetheless, there is still enough data to develop a general sense of the capital cost needed for a given level of plasma performance for each concept. This data is summarized in Table 9-2. Most of the data is from US machines, so cost differences due to accounting systems and labor costs are minimized. As subsequently used, these costs have been scaled to US 1990 dollars based on the Marshall & Swift annual equipment cost index [9.9].

The comparison of nxT performance against capital cost is shown in Figure 9-2. This figure is limited to achieved experimental results, or next-step designs (e.g. ITER, but not DEMO). The data does not strongly suggest that any one fusion approach provides better mT performance at lower cost than the others. However, the scatter is large and the real cost differences can be significant.

If we consider next-step fusion machines with significant fusion power (ignited or driven), then Table 9-2 summarizes the major characteristics of the most recent concepts. No conceptual designs of stellarator or compact toroid ignition experiments have been published recently. The construction costs are as reported by the study authors. The use of tritium, and significant neutron production, significantly affects the project costs.

9.2 Reactors

Although any given fusion concept may be able to achieve breakeven or ignition with a large enough experiment, this is not sufficient. The final product is a fusion power reactor, which must produce net power in a reliable, economic and safe manner. The concepts considered in the previous sections are advanced enough for at least preliminary considerations of a reactor. Table 9-4 summarizes several key parameters for recent representative designs. Although cost is used in most of these studies in order to determine the best design point, it is less reliable for cross-study comparisons and is not included here.

These designs assume some degree of extrapolation in engineering and physics. The extent of the physics extrapolation may be inferred from Table 9-1. The extent of the engineering risk can be inferred from such parameters as the neutron wall load and the peak magnetic field, from the relative size comparison figures in the appropriate sections of this report, and from the corresponding discussion of the key issues in this report. 54.

REFERENCES

9.1 US Congress, Office of Technology Assessment, 'STARPOWER, The US and the International Quest for Fusion Energy', OTA-E-338 (Washington DC, October 1987).

9.2 F. Ribe and D. Baldwin, "Fusion Program Planning for the Early to Mid 1990s', J. Fusion Energy, 7(4) (1988) 371.

9.3 R.L. Hagenson and R.A. Krakowski, 'Steady-State Spheromak Reactor Studies', Fusion Tech. 8 (1985) 1606.

9.4 F. Najmabadi, R.W. Conn, R.A. Krakowski et al, 'Engineering and Physics of High- Power-Density, Compact, Reversed-Field Pinch Fusion Reactors: The TITAN Study', Fusion Technology 1988, A.M. Van Ingen et al (eds.), Elsevier Science Publishers B.V. (1989) p.1779 (Proc 15th SOFT, Utrecht, 1988).

9.5 C.G. Bathke, R.A, Krakowski, R.L. Miller and K.A, Werley, The Reversed-Field-Pinch Fusion Neutron Source: A Conceptual Design', J. Fusion Energy 8,(3/4) (1980) 249.

9.6 R.A. Krakowski, J.D. Sethian and R.L. Hagenson, The High-Density Z-Pinch as a Pulsed Fusion Neutron Source for Fusion Nuclear Technology and Materials Testing', J. Fusion Energy 8(3/4) (1989) 269.

9.7 J.D. Lee (ed.), 'MINIMARS Concept Design: Final Report', UCID-20773 (Sept 1986).

9.8 R.L. Hagenson and R.A. Krakowski, 'A Compact-Toroid Fusion Reactor Based on the Field-Reversed Theta Pinch: Reactor Scaling and Optimization for CTOR', Proc 4th Topical Mtg on Tech. of Contr. Nuclear Fusion, King of Prussia, Pennsylvania, 1980 October 14-17, Vol.2 (1981) 1121.

9.9 Chemical Engineering, March 1990, 182.

9.10 C.C. Baker et al, Tokamak Power Systems Studies - FY 1985', Argonne National Laboratory, ANL/FPP/85-2 (December 1985).

9.11 R.L. Miller and the ARIES Team, The ARIES-I High-Field Tokamak Reactor: Design Point Determination and Parametric Studies' (IEEE 13th Symp. on Fusion Engineering, Knoxville, Tennessee, 1989 October 2-6) in The ARIES Tokamak Reactor Study', UCLA-PPG-1274 (October 1989).

9.12 Y-K.M. Peng, 'Initial Characterization of Spherical Torus Reactors for ARIES', FEDC- letter, M-89-ARIES-006,1989 August 23.

9.13 B.A. Carreras et al, 'Progress in Stellarator/Heliotron Research: 1981-1986', Nucl Fusion 28(9) (1988) 1613. 55. 9.14 R.A. Krakowski, J.D. Sethian and R.L Hagenson, 'The High-Density Z-Pinch as a Pulsed Fusion Neutron Source for Fusion Nuclear Technology and Materials Testing', J. Fusion Energy 8(3/4) (1989) 269.

9.15 C.C. Baker et al, 'An Overview of the STARFIRE Reference Commercial Tokamak Fusion Power Reactor Design', Proc. 4th Topical Mtg. on Tech. of Contrl Nuclear Fusion, King of Prussia, Pennsylvania, 1980 October 14-17, Vol.II (July 1981) 1074.

9.16 J.F. Lyon fed.), Special Issue on Stellarators, Fusion Tech. |7(1) (1990).

9.17 Y-K.M. Peng, 'Remarks', J. Fusion Energy, 8(1/2) (1989) 12.

9.18 J.-P. Watteau, 'Summaries of the 12th Inter. Conf. on Plasma Physics and Controlled Nuclear Fusion, Nice, France, 12-19 October 1988', Nucl. Fusion 29(3) (1989) 520.

9.19 D.B. Harris et al, 'Inertial Fusion in the Nineties', J. Fusion Energy 7(2/3) (1988) 163.

9.20 Panel Discussion, 'Progress in Inertial Fusion1, J. Fusion Energy £(1/2) (1989) 51.

9.21 Fusion Power Associates Newsletter, September 1990. 56.

Table 9-1. Best plasma performance for various fusion approaches.

Best Achieved nxT Projected Next-Step ntT*

Concept xETi(0) Machine Year TETj(0) Machine Startup OO^s-keV/m3) (10Ms-keV/m3)

Tokamak 4 JET 1988 30 CIT 1996 Laser 2 1988 - - - Tandem Mirror 0.04 GAMMA 1988 - - - Stellarator 0.02 W-VII-A 1986 0.6 LHD 1996 High-Density Z-pinch 0.002 HDZP 1988 10 HDZP.ZFX 1990 FRC 0.001 FRX-C 1989 0.006 LSX 1989 Reversed-Field Pinch 0.0002 ZT-40M 1984 0.003 RFX.ZTH 1991,1993 Spheromak 0.00004 S-1 1986 - - - Spherical Torus - - - - -

Machines under detailed design or construction. 57.

Table 9-2. Characteristics of Next-Step Ignited or Driven Machines

Concept Conventional High-B Stellarator Spherical Reversed Dense FRC Spheromak tokamak tokamak torus field Z pinch pinch

Facility CIT, Ignitex, - ORNL ZTI. FTF, - - PPPL U. Texas LANL LANL

Purpose Ignition Ignition - Ignition Ignition Neutrons - - Major radius, m 2.1 1.5 - 1.5 2.4 L=0.1 - - Avg minor 1 0.6 - 1.4 0.35 20 urn - - radius, m Magnet field, T 10 20 - 2 ~5 - - - Current, MA 11 12 - 14 9 1.9 - - Power, MWfc,,^ 300 150 - 60 124 8 - - Const, cost, 300-500 140 - 300 350 110 - - M$US90 58.

Table 9-3. Capital cost data for various existing and proposed machines*

Experiment Capital Cost Comments

Tokamaks OKI, General Atomic 110M$US87 Construction cost including JAERI contribution [9.1] ISX-B. Oak Ridge 5MSUS87 Construction cost [9.1] Alcator-C, MIT 15 M$USB7 Construction cost [9.1] TEXT, U. Texas 21 M$US87 Construction cost [9.1] PDX, Princeton 54 M$US87 Construction cost [9.1] PLT, Princeton 43 M$US87 Construction cost [9.1] TFTR, Princeton 600 M$US87 Construction cost [9.1] JT-60, Japan 700 M$US87 JET, Culham 900 M$US87 Construction cost [9.1] TdeV, Montreal 20 M$US87 FTU, Frascati 33 MSUS84 CIT, Princeton 500M$US86 Construction cost including site credits STE, US 465 MJUS90 Construction cost at new site ITER 5000 MSUS9O Construction cost SRX, PPPL/Grumman 12M$US89 Construction cost [9.2] Ignitex, U. Texas 130 M$US89 Construction cost

Stellarator and Reversed Field Pinch ATF, Oak Ridge 21 MSUS87 Construction cost [9.1] IMS, Wisconsin 1.4MSUS87 Construction cost [9.1] TJ-II, Madrid 40 M$US90 LHD. Toki 500 M$US89 [9.16] ATF-II, Oak Ridge 50-100 M$US89 [9.16] OHTE, General Atomic 20 M$US89 ZT-40M, Los Alamos 20 MSUS87 [91] MST. Wisconsin 3-10 M$US89 Project cost ZTH. Los Alamos 75 MSUS88 Construction cost RFX, Padua 90 M$US90 Project cost FTF/RFP, Los Alamos 335 M$US88 Project cost excluding site [9.5]

Other magnetic confinement concepts Proto-ETA, Brazil 3-10 M$US89 Project cost at green-field site ST-Steady-state, ORNL 20-40 MSUS89 [917| ST-Neutron source, ORNL 300 M$US89 [9.17] HDZP, Los Alamos 0.5 M$USB8 Device and power supplies [9.6] FTF/DZP, Los Alamos 110M$US88 Project cost excluding site (7 M$ for device/power supplies) [9.6] FRX-C, Los Alamos 3 M$US87 Construction cost (9.1] LSX, Spectra Tech 15M$US88 Construction cost [9.21] S-1, Princeton 8 M$USS3 MS. Maryland 1.5MSUS85 EBT-P, Oak Ridge 66 M$US82 Construction cost TMX-U, Livermore 30 M$US87 Construction cost, including TMX credits [9.1] TARA, Boston 19 M$US87 Construction cost [9.1] MFTF-B, Uvermore 200 M$US82 Construction cost MFTF-U, Livermore 400 M$US85 LLL-TDF, Livermore 1000 MSUS85 Project cost PHAEDRUS-Mirror 1.8 M$US87 Construction cost [9.1] Tokapole, Wisconsin 0.6 M$US87 Construction cost [9.1]

Inertial confinement NOVA, Livermore 200 M$US87 [9.19] PBFA-II, Sandia 48 M$US87 [9.19,9.20] Aurora, Los Alamos 40 M$US87 [9.191 Athena, Livermore 750 M$US88 [9.181

* Note that costs are not consistent. All include the basic device construction cost, but (unless labelled project cost) may not fully include auxiliary heating, power supplies, diagnostics, utilities, building, site and project engineering. 59.

Table 9-4. Representative DT Reactor design studies

Conv. High-B SSR Stall- Spher. RFP FRC Spher- Dense Tandem tokamak tokamak tokamak arator torus omak Z pinch mirror

Study Starfire ARIES-I TPSS ATR-2 ORNL TITAN CTOR LANL NRL MINIMARS Study date 1980 1989 1987 1989 1989 1987 1980 1985 1989 1986 Reference [9.15] [9.11] [9.10] [9-13] [9.12] [94] [9.8] [9.3] [9.14] [9.7]

Major radius, m 7.0 6.5 5.2 10.5 1.5 3.9 40* 2.2* NA 138* Avg minor radius, m 2.4 1.9 1.3 2.26 1.5 0.6 1.2* 1.7' 1* 0.4* Magnetic field on-axis,T 5.8 13 3.8 5.0 0.8 (8) 4.2 13.9 - 3.1 Plasma current, MA 10.1 12 4.0 -0 28 17.8 - 30 4 0 Pulse length, s SS SS SS SS SS SS 2 SS 40 Hz SS Average plasma beta. % 6.7 2 25 6.3 70 22 80 (H) (100) 60 Maximum field at coil, T 11 24 6 - (6) - - 3 NA 24 Engineering beta, % (2) (0.6) (10) - 6.5 - - (75) NA - Fusion power, MWf 3500 1991 1275 3500 2000 2290 840 770 346 1231 Net electric power, MWe 1200 1000 540 1200 980 900 310 250 100 600 Retire, power, %net 20 30 30 - . 14 18 36 (300)+ 30 Wall load, MW/m2 3.6 2.8 3.4 2.7 12 18 2.0 17.4 17 3.3 Pwr density, kWe/Mg^ 50 90 130 - 500 800 - 630 - 100 Magnetic energy, MJ/MWe 50 160 10 (20) (0.5) 2 (10) 1.4 (0.08) (16) Plasma energy, MJ/MWe (0.6) (0.6) (0.5) (1) (0.06) (0.1) (0.3) (0.2) (0.004) (0.2)

* Chamber dimensions, axial length is reported rather than major radius. () denotes estimated from other data in original reference. + Stored inductive energy which is directly recovered after pulse with high efficiency. TFTR

JET

'YIIB 1 0 a GEKKO-XII m JET •

NC I/A GAMM; 10 • , FRX-C/ a • Comp' Ti(O) (keV) i W-VII- A O«* A S;1 TRX-2* • Heliotron-E i J « ZT40M ^ O

A A J H3ZP < OKIE 1 ,

1

1 1.00E+15 1.00E+16 1.00E+17 1.00E+18 1.00E+19 1.00E+20 1 00E+21 1.00E+22 -tau (s/m3)

Tokamak O Stellarator * RFP Mirror A CT Pinch " Laser

Figure 9-1. Lawson diagram for various fusion approaches, (data achieved experimentally as of 1989). 1.00E+24

1 00E+22

s

• 1.00E+20 o o • •

• X • n-tau-T 1.00E+18 (s-keV/m3) A • « A o A s 1 00E+16 A * a o s a a 1.00E+14

D 1 00E+12 10 100 1000 10000 Capital Cost (M$US90)

• Tokamak © Stellarator a RFP Mirror A ST * CT X Pinch - Laser

Figure 9-2. Achieved and projected (next-step machines) physics performance as a function of actual or estimated construction cost. 62.

APPENDIX A: OTHER CONCEPTS

A variety of other fusion concepts have been proposed. Most of these have a particularly limited database, or an unresolved major issue. The nature and status of several such concepts are summarized here: - Colliding beam fusion (migma) - Electrostatic confinement - Muon-catalyzed fusion - Spherical pinch - - Linear systems - Miscellaneous concepts

A.1 COLLIDING BEAM FUSION (MIGMA)

Simple energy considerations specify that less than 1 MeV of kinetic energy are required for hydrogen to overcome their mutual Coulomb repulsion and, upon fusion, yield -4-20 MeV. The simplest means to obtain such energy multiplication would be to direct the beams from two ion accelerators towards each other for direct ion-ion collisions. Early research, however, quickly established that ion-ion scattering, with the consequent loss of fuel ion energy and direction, rendered this approach energetically futile.

In the late 1960's, B. Maglich suggested imposing an external magnetic field upon a colliding beam reaction volume such that the scattered ions would be returned repeatedly to the reaction region until they fused. The concept, called "Migma fusion", involves injecting medium-energy ions into a disk-shaped reaction chamber placed between a pair of solenoidal coils. The JxB magnetic force on the ions induces large gyroradius precessional orbits leading to high probability of near-direct collisions in the geometric center.

The magnetic field topology and operational features suggest that the Migma concept resembles a short-axis continuously driven by accelerator injected fuel ions. However, the plasma consists of a highly ordered ion population in a non-equilibrium non- Maxwellian state. Global charge neutrality is sustained by an ambient population of oscillating electrons. At sufficient central ion density, a diamagnetic field is generated which may locally cancel the external magnetic field.

A series of experiments were conducted until the mid-1980's at the Aneutronic Energy + Laboratory, Princeton. The last experiments (Migma IV) involved 0.5 mA of D2 accelerated to 0.7 MeV. Selected results are given in Table A-1 [A.1 ,A.2]. Note that the normal numerical value does not apply to this non-thermal system.

The most attractive features of the concept are its simple geometry, potential suitability for advanced fuels (the plasma is non-thermal), and potential useful power output in the 10-100 MW range. A critical technical issue relates to the central axis ion density exceeding limits (e.g., space charge, or ). Other questions concern confinement and 63. The concept received some US governement support, but no US DOE funds. Only a minimum of research is now underway. Selected theoretical support has been provided by small groups elsewhere.

A.2 ELECTROSTATIC CONFINEMENT

The concept of electrostatic confinement emerged from an observation in the 1930's that a special "glow" appeared in the center of some spherically symmetric vacuum tubes. It was reasoned that this was due to the formation of a space-charge potential which trapped ions. In the mid-1950's, P.T. Farnsworth reinterpreted the central glow and suggested it could be applied to fusion. Systematic research on electrostatic confinement for fusion applications began in the mid-1960's starting with the ITT Farnsworth Research Corporation.

In the electrostatic confinement concept, a spherical anode emits ions (eg., D+) which are accelerated towards the center. These ions pass through a negatively charged grid which is about 95% transparent to ions. As the ions accumulate in the center, a positive space charge builds up tending to increasingly reverse the motion of the ions towards the outer anode - only to be reversed by electrostatic repulsion by the real anode towards the centre. An important consequence of this repeated ion motion reversal is the formation of a virtual anode defined by the existence of a high density ion shell inside the cathode. The inner cathode grid emits electrons only toward the center and not to the outer real anode. As the electrons are accelerated towards the center, they pass through the virtual anode ~ note that the medium is a globally neutral plasma - to form a negative space charge tending to, as for the ions, reverse the electron motion and by a similar process form a virtual cathode. If the ion density and temperature is sufficiently high, then non-Maxwellian, beam-type fusion reactions will occur.

Electrostatic confinement was of experimental interest until the mid-1970's [A.3-5]. There was considerable debate as to whether the neutrons were from beam-background or beam-beam reactions. This could not be resolved by the available diagnostics and interest declined. Still, the research led to the following results: - Potential well formation and ion trapping is readily established; - Ion densities of up to 1019/m3 and neutron production rates of up to 1010 n/s; - Some design variations (eg. cylindrical geometry) and use of small accelerators for ion injection appear readily adaptable.

No fusion reactor design exists, although some concepts have been proposed. In principle, the reactor could be very compact. on the real anode and cathode would be a major concern.

Current research is being undertaken in the US (U of Illinois [A.21], Columbia U., Directed Technologies Inc.) and in the USSR (Kharkov). Recently, a large contract has been issued by the US DARPA agency to Directed Technologies to construct a proof-of-principle experiment using magnetic field confinement of the electrons rather than grids. 64. A.3 MUON-CATALYZED FUSION

The possibility of muon-catalyzed fusion was first noted in the 1940's, but appeared to have serious physical limitations that made it unattractive. However, in the 1970's a resonance reaction was found that increased the reaction rates considerably and led to renewed interest.

In muon-fusion, the electron in a DT molecule is replaced with a negative meson, called a muon, which is 207 times heavier than the electron. The resulting molecule is physically smaller (by a similar factor), and so the fusion probability between the two nuclei is increased by many orders of magnitude. The muon is not consumed in the fusion reaction. Rather, it is released and may be available to catalyze further DT molecules into fusing.

The advantages of muon-catalyzed fusion are that the reaction occurs in an ordinary gas, rather than a plasma. No magnetic fields are necessary, and the temperatures and are within the range of normal engineering. However, the muon has a lifetime of 2.2 us for radioactive decay, and requires about ~4 GeV to produce (e.g., in accelerators). Therefore, practical net energy gain requires that each muon must catalyze > 1000 DT reactions before decaying. This is critically dependent on whether the muon is captured by the fusion alpha particle product [A.6],

Present experiments at up to 900 K and 10 MPa have achieved -170 catalyzed DT reactions per muon. At this rate, muon-catalysis may be of interest for fission-fuel production, but not for fusion power. However, three ideas have been suggested for fusion power. In one, muons are formed in-situ in a magnetically-confined plasma (for more efficient muon formation and retention). In the second, muons are directed into a very high density medium such as that in an inertial confinement target or a dense-Z pinch (to reduce the alpha capture cross- section). And finally, an intense laser beam might be used to strip muons from the alphas.

A.4 SPHERICAL PINCH

In the classic Z- and 0-pinches, a plasma is formed near the cylindrical chamber walls and compressed inwards by either axial (©-pinch) or azimuthal (Z-pinch) magnetic fields. These were found to have unacceptably short stability and/or high end losses.

The spherical pinch concept was developed by E. Panarella in the 1980's at the National Research Council laboratories (Canada). The main features are the use of an imploding spherical Shockwave to compress, confine, and heat a small plasma formed at the machine center.

Early experiments used a slotted metal shell to force the current to flow uniformly on a sphere and inductively create a spherical Shockwave, which converged onto a central spark discharge (Figure A-1). For fusion relevant conditions, scaling arguments [A.7] indicated -1 MPa gas fill pressures, with high power density deposited to form the central plasma and shock waves. Inductive discharges cannot operate at such pressures. Therefore, the last series of experiments used resistive (spark) discharges in a cylindrical geometry. The experimental 65. results are summarized in Table A-1. For a capacitor bank energy of 1 kJ, about 107 neutrons/pulse were observed with an inferred nx of 2 x 1020 s/m3 and ion temperature of 0.7 keV [A.8].

The most critical issue is to test the Shockwave formation and compression in larger geometries with greater power input. It has been proposed that the central plasma be formed using a laser pulse, and the spherical shock waves produced by a set of paired-electrode discharges situated around the circumference of the spherical chamber. The protection of these electrodes, and possibly the power transfer interface into the vacuum chamber, are significant engineering questions. There has been no reactor design study of a spherical pinch.

At present, a limited amount of research is underway, primarily for generating X-rays for lithography, although there are proposals for larger fusion-relevant experiments.

A.5 DENSE PLASMA FOCUS

The plasma focus is based on the principle that an electrical discharge in a gas will form a plasma sheath which, by device design, may accelerate towards a focal point thereby producing a high ion density plasma in a small volume. The remainder of the current then discharges through the plasma and heats it. Such devices possess an inner cathode and outer anode in either cylindrical or disk-shape geometry. The associated dynamics in the region of collapse renders the plasma focus system essentially a Z-pinch pulsed device and is schematically described in Figure A-2 [A.9-11].

The plasma focus was invented in the 1960's and for the next 15-20 years proved to be a widely used device for the study of high density plasmas. The low cost of manufacture, reliability of operation, and the availability of fast optical/electronic diagnostics accounted for its wide use. The neutron production per pulse was generally taken as the most relevant indicator of performance and thereby suggested its potential utility for fusion energy purposes.

The typical parameters of the more powerful installations are the following [A. 10]: Capacitor bank energy 500 kJ Capacitor bank voltage 100 kV Compression time 1 us

Plasma Foci have proven to be operationally reliable, but exhibit complex plasma instabilities. All experiments designed to relate neutron yield to variation of some experimental parameter have found limits. It is believed that unavoidable plasma impurities from the insulators are responsible for these upper bounds. It is also now thought that beam-target interactions within the focus, rather than thermal reactions, are responsible for much of the neutron output. Therefore, dense plasma foci are not considered suitable for fusion power reactors. However, their large neutron yield for modest input energy has suggested other applications - for example, some modest scale-up and use of parallel plasma focus devices attached to a large common fissile-fertile blanket [A.11]; or as X-ray sources. 66. Interest in plasma focus research has declined in recent years. The more active groups are: - Los Alamos National Laboratory, New Mexico, USA (A.H. Glasser) - P.N. Lebedev Physical Institute, Moscow, USSR (V. Gribos) - Institute of Plasma Research, University of Stuttgart, FRG (H. Herold, H. Schmidt) - Institute of Nuclear Studies, Swierk, Poland (A. Jerzykiewicz)

A.6 LINEAR SYSTEMS

Linear magnetic fusion systems were among the first concepts to be tested because of their apparent simplicity. By the late 70's and early 80's, these linear systems had generally evolved into concepts that were either high-density, short-lived and fairly compact, or low- density, long-lived and physically long [A.12, A.13].

The Z-pinch evolved from a long cylinder of gas to a short (5 cm) wire of frozen D2 which is heated to ignition by an axial current. The reaction is over too quickly for end losses to matter. As such, it has little in common with the other linear magnetic systems described here, and is treated separately in Section 6.

In the LINUS approach, a conducting tube is filled with a weak field and a cold plasma. The tube is then compressed by electromagnetic forces, and the trapped plasma is heated by compression. First experiments used aluminum tubes which were destroyed with each pulse, but reactors would use liquid . The liquid lithium liner would be stabilized by rotation, and pushed back by the expanding plasma, leading to direct energy recovery. The key issues were reducing end losses, and the stability of the liner during implosion and recovery. [A.12.A.14]

The laser-heated solenoid uses high-efficiency CO2 lasers to heat a plasma in a solenoidal magnetic field to ignition (Figure A-3). The end losses are high, so the reactor must be 400- 1000 m long. However, the simple linear geometry should make engineering straightforward. Nevertheless, besides physics uncertainties in the trapping efficiency of the laser beam and the stability of the plasma, very high magnetic fields of 30-50 T are required to confine the high-density plasma (lO23"24^3) needed to keep the reactor length reasonable [A.12]. And even under these conditions, the long coolant piping lengths and large number of modules (and connections) largely countered the possible advantages of linear simplicity [A.13].

The electron-beam heated solenoid replaces the laser-heated solenoid with an electron beam heating source. This should be much more efficient in beam generation (75%, versus 25% for the CO2 laser), and possibly more efficient in absorption by the hot plasma at lower densities (<10Z2/m3). The beam focusing and absorption processes are, however, not as clear as with lasers and the problems of end-losses and long length remain. [A.12]

Linear mirrors offer the advantages of high beta operation, simple geometry, a natural divertor, and close agreement between theory and experiment. Experiments have achieved quite reasonable plasma parameters. As methods for reducing the end losses became more sophisticated, the original simple mirror concept evolved into a tandem mirror, which became 67. a tandem mirror with thermal barriers. In the tandem mirror, a central "conventional" simple mirror is connected at each end with another mirror. By driving the plasma density very high in these end mirrors using neutral beams, it is possible to create an electrostatic barrier that enhances the ion confinement in the central cell beyond that due to the magnetic mirror alone. By making the power-producing central mirror sufficiently long, one could always get enough power to counter that lost out the end mirrors. In fact, it was still found necessary to separate the electron population in the central and end mirrors by various 'thermal barrier' concepts, and to use direct energy recovery on the escaping plasma. These end-stoppering regions were becoming quite complex, and the reactors were still relatively long and low-power density. At present, only Japan and the USSR have active mirror research programs.

Ultimately, in addition to the specific physics and engineering uncertainties for each of the linear magnetic concepts, the common problem of end-losses has remained intractable. Work on linear systems for fusion reactors has largely stopped.

A.7 MISCELLANEOUS CONCEPTS

An Ultra-Low-Q (ULQ) plasma is a toroidal plasma similar to a reversed-field pinch, but without field reversal. Like a reversed-field pinch, the plasma shows considerable self-organization. In particular, it maintains particular values of the current (corresponding to near-whole-fraction values of q) despite changes in the toroidal loop voltage driving the current. The 'safety factor', q, is the magnetic field pitch defined as the number of toroidal turns of the magnetic field per poloidal turn. For tokamaks, q must be above unity everywhere for stability, while for reversed-field pinches it varies from positive to negative across the plasma radius. For ULQ discharges, 0 < q < 1. In experiments on OHTE and TORIUT, for example, ULQ behavior has been seen at 140 kA currents and 0.4 T toroidal fields, for globally stable discharges of over 10 ms, with Ti ~ 0.3 keV and nx ~ 1016 s/m3 [A.15,A.16]. MHD instabilities has largely led to lack of interest in this approach for fusion.

The is a series of magnetic mirrors arranged into a torus to avoid end losses. In order to provide stability to compensate for that lost due to the toroidal curvature, the plasma within each mirror is encircled by a hot electron ring. Early experiments on ELMO (Oak Ridge), EBT-I (Oak Ridge), and NBT-I (Nagoya) were successful, and the concept was proposed as the major alternate fusion concept within the US. Reactor studies indicated high beta, steady-state plasmas, but with high aspect ratios. Subsequently, operation of EBT-S found a class of poorly confined particles that effectively resulted in the same end loss problem as with linear mirrors. Work on bumpy tori was stopped in the US, while NBT-1M was built in Japan and operated as a physics experiment.

Multipoles create cusp-like magnetic fields on the plasma surface that confine the plasma in a central low-field region (a magnetic well). The advantages are high beta (due to the low field), steady-state operation, and natural divertors [A.12]. Various configurations have been tested, from toroidal devices with 4-6 pairs of conductors (e.g., Octupole at GA, Tormac at U. Wisconsin, Dodecapole at UCLA), to linear mirror-based devices with three pairs of conductors (Linear Surmac, Figure A-4). In multipoles, the conductors are surrounded by plasma, so 68. specific issues are how to support and protect the conductor within the plasma, and the high recirculating power needed for copper conductors as superconductors cannot be used. In addition, there are the issues of stability and cusp-losses for toroidal devices and end-losses for linear devices.

The Extrap concept stabilizes a Z-pinch using an external poloidal field (Figure A-5). The pure poloidal field decreases to zero on axis, which allows 6-1 Axial losses should be slower in a linear Extrap since there is no axial field. The limited experiments confirm that linear Z- pinch stability is improved by adding an octupole field, although present experiments may be stabilized by finite Larmor radius effects (the ratio of plasma minor radius to ion gyroradius is ~3-8, compared with 30-50 in a reactor). First experiments with toroidal devices have been found to be less stable than with a linear device. [A.16,A.18,A.19]

The use of ion- or electron-ring plasmas has been proposed for a variety of configurations, e.g. , and there are several small experiments. They have often been considered as a means to stabilize other confinement concepts, such as the electron rings stabilized a toroidally-bent set of mirrors in the Elmo Bumpy Torus concept, or in recent proposals to use rings to stabilize Field-Reversed Configurations.

The Plasmak is a spheromak torus stabilized by a surrounding spherical plasma rather than a metal shell. This plasma mantle, possibly a relativistic electron ring, is in turn surrounded and stabilized by neutral gas [A.14, A.20]. (See Figure A-6).

REFERENCES

A.1 B.C. Maglich, "Time Average Neutralized Migma: A Colliding Beam/Plasma Hybrid Physical State as Aneutronic Energy Source", Nucl. Inst. Meth. A271 (1988) 13.

A.2 Proc. 2nd Inter. Symp. on Aneutronic Power, 1989 April 28-29.

A.3 R.G. Hirsch, 'Inertial-Electrostatic Confinement of lo.iized Fusion Gases', J. Appl. Phys. 38(1967)4522.

A.4 T.J. Dolan et al, 'Electrostatic-Inertial Plasma Confinement1, J. Appl. Phys. 43 (1972) 1590.

A.5 W.M. Black and E.H. Klevans, Theory of Potential-Well Formation in an Electrostatic Confinement Device', J. Appl. Phys. 45 (1974) 2502.

A.6 Yu.V. Petrov, 'Muon Catalysis for Energy Production by Nuclear Fusion', Nature 285 (1980)466.

A.7 E. Panarella and P. Savic, 'Scaling Laws for Spherical Pinch Experiment! , J. Fusion Energy 3 (1983) 199. 69. A.8 E. Panarella, The Spherical Pinch, J. Fusion Energy 6 (1987) 285.

A.9 A.H. Glaser, HZ-Pinch and Plasma Focus", Nuclear Fusion 29 (1989) 129.

A.10 H. Herald et al, "Comparative Analysis of Large Plasma Focus Experiments Performed at IPD, Stuttgart, and IPJ, Swierk", Nuclear Fusion 29 (1989) 1255.

A.11 A.A. Harms and M. Heindler, "The Matching of Dense Plasma Focus Devices with Fission Reactors", Nuclear Science Eng. 66 (1978)

A.12 F.F. Chen, 'Alternate Concepts in Controlled Fusion, Part A, Executive Summaries', EPRI ER-429-SR, May 1977.

A.13 N.A. Krall, 'Alternate Fusion Concepts as Reactors', in 'Unconventional Approaches to Fusion', B. Brunelli and G.Leotta (eds.), Plenum Press (1982) New York.

A.14 V.E. Haloulakos and R.F. Bourke, 'Fusion Propulsion Study', AL-TR-89-005, Air Force Space Technology Center, July 1989.

A.15 T. Tamano et al, 'Studies of Plasma Self-Organization in Toroidal Pinches', in Physics of Mirrors, Reversed Field Pinches and Compact Tori, S. Ortolani and E. Sindoni (eds.), Vol. I (1988) 145.

A.16 J. Sheffield, 'Summaries of the 12th Inter. Conf. on Plasma Physics and Controlled Nuclear Fusion, Nice, France, 12-19 October 1988', Nucl. Fusion 29(3) (1989) 499.

A.17 'World Survey of Activities in Controlled Fusion Research, 1986 Edition', IAEA Vienna (1986).

A.18 J.E. Eninger and B. Lehnert, The EXTRAP Fusion Reactor Concept', Fusion Technology 1988, A.M. Van Ingen et al (eds.), Elsevier Science Publishers B.V. (1989) (Proc. 15th SOFT, Utrecht, Netherlands) p.1799.

A.19 B. Lehnert, The Extrap Concept', in 'Unconventional Approaches to Fusion', B. Brunelli and G. Leotta (eds.), Plenum Press (1982) New York.

A.20 P.M. Koloc, 'Plasmak™ Power for Energy Intensive Space Applications', Fusion Tech. 15 (1989) 1136.

A.21 G.H. Miley et al, 'Inertial-Electrostatic Confinement: An Approach to Burning Advanced Fuels', Proc. 9th Top. Mtg. on Tech. of Fusion Energy, Oak Brook USA, 1990 October 7-11. 70.

Table A-1. Parameters of various alternative fusion experiments

Facility/Location Year LorR T,(0) Pu Comments/Status o 1 3 (m) (m) (T) (MA) (s) (1/m ) (s/m ) (keV) (MW) Focussed Plasma Migma-IV, US 1982 - 009 3.5 - - 3ei6 1e18 300* - Colliding beam [A.1.A.2] Spherical pinch, Ottawa 1985 - 0.02 - - - - 3e25 1e20 1.3 Spherical pinch [A.7.A.8] KPF, USSR 1986 - - 3 - - 3e24 - 0.5 - Plasma focus, 1e11 n/shot |A 17) Posekten, Stuttgart 1986 - - - 2.4 - - 8e24 - 1 Plasma focus, 2e11 n/shot (A. 17) PF 360, Swierk 1986 - 2.5 - - 1e25 - - - Plasma focus. 2e11 n/shot [A. 17] SHUDEN. Osaka 1988 - • - 1.2 - - 1e23 1815 11 - Plasma focus, 5eB n/shot (A.17|

Toroidal Plasma Scyllac, Los Alamos 1968 ------2e17 5 - Toroidal theta pinch EBT-S. Oak Ridge 1982 1.5 0.16 1.0 - - - 1e18 1e16 0.01 0.3 Bumpy torus [A. 17] NBT-1, Nagoya 1982 1.6 0.2 0.3 - - - 1e18 1e15 0.05 - Bumpy torus [A. 17] Dodecopole, Los Angeles 1983 0.4 0.1 0.2 - 8 3e19 0.2 - Toroidal multipole Octupole. Wisconsin 1983 1.4 0.25 0.1 - 2 1e19 - 0.02 - Toroidal multipole TCX-I, Los Angeles 1986 0 45 0.15 0.006 - - 5e19 2e15 0.5 - Toroidal multipole (A. 17) SPICA, Netherlands 1986 0.45 0.5 1.3 0.5 0.2 8 3e21 - 0.05 - Toroidal screw pinch (A. 17) NBT-1M. Nagoya 1986 1.4 0.1 1 - 0.01 4 1e19 1e14 0.1 0.3 Bumpy torus [A. 17) EXTRAP-T, Sweden 1988 0.45 0.04 Q.\ 0.03 0.05 50 8e20 1e16 0.03 . Toroidal Z-pinch IA.17.A.18] NBT-iM, Nagoya - 1.4 0.14 1.0 - - - 4e18 2e16 0.1 - Bumpy torus (A. 17] EBT-P, Oak Ridge - 4.5 0.3 2.1 - - - 1e19 5e17 0.4 1.2 Bumpy torus, 1982 proposal

Linear Plasma 2XIIB, livermore 1978 - 1e20 3e16 8 5 Single mirror PHAEORUS. Wisconsin 1986 3 0.19 0.06 0 0.03 4 3e17 3e13 0.1 - Tandem mirror [A. 17) TMX-U. Livermore 1986 8 0.25 0.3 0 0.1 Ie18 5e15 1.5 1.5 Tandem mirror [A. 17) KP, USSR 1986 2 0.05 10 0.01 0.0002 70 5e20 1e15 2 - Linear theta pinch (A. 17] TARA, Boston 1988 10 0.2 0.22 0 0.1 2e18 1e15 0.3 1 Tandem mirror (A. 17) GAMMA-10, Tsukuba 19b i 5.6 0.18 0.64 0 0.02 2e18 1.8 0.2 Tandem mirror [A. 16.A. 17] MFTF-B, Livermore - 20 0.45 1 0 0.03 3e19 3e15 10 2 Tandem mirror (A 17| MFTF-U, Livermore o Tandem mirror. 1985 concept LLL-TDF, Livermore - - 0.2 4.5 0 6e20 7e18 30 70 Tandem mirror, 1985 concept

* Non-maxwellian energy SPHERICAL METAL VESSEL ('SPHERICAL PINCH')

EXPANDING CONTACT INDUCTIVE SURFACE DISCHARGE

CF .TRAL DISCHARGE IMPLODING SHOCKS

POLE

Figure A-1. (a) Schematic diagram of spherical pinch chamber; (b) Slotted shell of chamber illustrating current paths used to inductively form spherical shock waves [A.8]. PLASMA SHEATH ACCELERATION

INSULATOR

CAPACITOR POWER . SUPPLY

SHEATH COLLAPSE AND FORMATION OF HIGH-DENSITY PLASMA

Figure A-2. Schematic diagram of Plasma Focus illustrating plasma sheath acceleration and focussing stages.

Superconducting Solenoid'

Conventional Solenoidal Magnet

Plasma

Blanket

Incident Laser Beam

Figure A-3. Schematic diagram of Laser-Heated Solenoid reactor [A.12]. Rtturn Conductor!

Critical Flux Surf

Multipolt Conductors

Figure A-4. Linear Surmac with six conductors along the plasma surface. The vacuum chamber would be near the critical flux surface contour. [A.12]

Figure A-5. Magnetic field lines in a toroidal Extrap [A. 19]. &v ' MANTLE •

KERNEL - GAS BLANKET

•. • •

VACUUM ' :• v PLASMA MAGNETIC. ' * FIELD • * m •

Figure A-6. Schematic diagram of the Plasmak™ concept [A.20]. 75.

APPENDIX B: INERTIAL CONFINEMENT

INTRODUCTION

Inertial fusion includes laser and ion-beam approaches. The laser program is focussed primarily on glass () lasers and gas (KrF) lasers; the former because they work and the latter because they may ultimately be more energy efficient. The ion-beam fusion is a smaller effort overall, and is considering light ions (Li+) and heavy ions. Each of these approaches requires considerably different driver technology.

Within the last decade, considerable progress has been made on target issues:

1) It has become clear that laser wavelengths on the order of 0.3 urn provide the best combination of laser-target coupling and practical technology. This has resulted in the closing of large CO2 (10.6 um) laser programs, development of frequency tripling for conventional Nd- glass lasers (from 1.06 um to 0.35 urn), and development of alternate lasers which lase at the appropriate wavelengths (e.g., 0.25 um KrF).

2) The problem of maintaining spherical symmetry during the implosion is being addressed by indirect drive using 'holrauhm1 targets, and by direct drive. In indirect drive, the target is surrounded by an outer shell which, upon irradiation by a few beams, produces a cascade of primary and secondary X-rays which produce an overall symmetric driving flux on the surface of the inner fusion target. The feasibility of this method has apparently been demonstrated in a series of classified nuclear tests called Centurion/Halite. The direct drive approach uses multiple beams to provide a direct, uniform irradiation and compression of the target. Various beam smoothing approaches (either spatial or frequency) are needed to smooth out irregularities that would otherwise prevent uniform irradiation. Indirect drive requires more complex (and classified) target design and also more initial laser power. Direct drive requires more laser beams.

3) The development of low-density foam targets allows the use of thick, free-standing liquid DT layers, which are desirable for high-gain. These targets appear to be relatively easy to manufacture in quantity, making reactor-target manufacturing more practical. [B.2]

The next-step experiment requires delivering about 10 MJ onto the target. This step is referred to in the US as the Laboratory Microfusion Facility, the rough equivalent of CIT or ITER in the magnetic confinement fusion. Achieving this power at acceptable cost is a key issue for all drivers. For solid-state lasers, the key issues are achieving high efficiency and high repetition rates (heat transfer limits). For gas lasers, the amplifier modules require greater scale-up (factor of 2000 for KrF) than glass lasers (factor of 200 for 0.35 um Nd), and the projected high efficiency (5-10%) and high repetition rates must be demonstrated. For light ions, the major issues are focussing and pulse shaping at sufficient diode standoff distances. For heavy-ion drivers, the major specific issue is transport of the ion beam through the reactor cavity, and scaling in general as this has been the least explored option to date. [B.2] 76. Several of the major inertiai lob'on experiments are summarized in Table B-1.

REFERENCES

B.1 J.-P. Watteau, 'Summaries of the 12th Inter. Conf. on Plasma Physics and Controlled Nuclear Fusion, Nice, France, 12-19 October 1988', Nucl. Fusion 29(3) (1989) 520.

B.2 D.B. Harris et al, 'Inertiai Fusion in the Nineties', JrnI Fusion Energy 7(2/3) (1988) 163.

B.3 Panel Discussion, 'Progress in Inertiai Fusion', JrnI Fusion Energy 8_(1/2) (1989) 51.

B.4 K. Imasaki et al, 'LIB-ICF Studies at ILE, Osaka U.\ in Laser Interaction and Related Plasma Phenomena, Vol.8 (H. Hora and G.H. Miley, eds.), Plenum Press, New York (1988) p.633.

B.5 J. Ramirez et al, 'Performance of the Hermes-Ill Gamma Ray Simulator', Proc. 7th IEEE Pulsed Power Conf., Monterey, California, 1989, p.26. 77.

Table B-1. Parameters of major inertial fusion experiment*

Facility/Location Year E Beam pR tpu|M iE ^(0) Comments/Status (kJ) type (mg/cm*) (W/cm2) (mm) (ns) (g/cm3) (s/m3) (keV)

Present/Completed JANUS. US 1974 0.01 1.06 urn Nd - 0.1 3e12 0.5 ARGUS, Livermore 1978 2 1.06 urn Nd 0.3 3e19 0.5 1e9 n/shot SHIVA, US 1979 3 1.06 urn Nd 0.3 3e19 0.5 20 beam, 3e10 n/shot Asterix, France 1983 0.3 1.3 urn 1 0.3 . 1 beam

ANTARES, Los Alamos 1985 40 10.6 urn CO2 1 1e10 n/shot GEKKO-XII. Osaka 1986 . Nd . 4e18 7 [B.1] Chroma, KMS Fusion 1988 1 Nd - 2 TW [B.3] NOVA, Livermore 1988 17 0.35 urn Nd 2 20 1e20 2 10 beam, 1e13 n/shot [B.1.B.2] OMEGA. Rochester 1988 1.5 0.35 urn Nd 30 0.25 1 - 1 [B.1,B.3] ASHURA, Tsukuba 1988 0.71 0.25 urn Kr 1e12 - 95 - -. [B.1] PBFA-II. Sandia 1988 75 H+ 10 <1 - 36 beams [B.1,B.2.B.3] REIOEN-IV, Japan 1988 0.4 3 MeV H+ 1e11 5 20 [B.4] Hermes-Ill, Sandia 1988 . e - 600 40 - [B.5] Aurora, Los Alamos 5 0.25 urn Kr 5 48 beam, under const [B.1 ,B.2] Super Sprite, UK - 3.5 0.25 urn Kr 1 - under const [B.1] Nike, NRL - 3 0.25 urn Kr 5 under const [B.1] GSI, Darmstadt - - heavy-ion - 0.1 70 - 0.02 under const [B.3]

Planned/Proposed GEKKO upgrade, Osaka - 100 0.35 urn Nd 300 2 5 24 beams, 1989 proposal [B.1] OMEGA upgrade, Roch - 30 0.35 urn Nd 300 5 60 beams, 1989 proposal (B.1] Athena, Livermore - 164 0.35 urn Nd . . 1988 concept [B.1]

* Simultaneous values (some inferred). E is beam energy brought to target, not necessarily coupled into target.

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