Initial Exploration of High-Field Pulsed Stellarator Approach to Ignition Experiments

In the framework of fusion energy research based on magnetic confinement, pulsed high-field tokamaks such as Alcator and FTU have made significant scientific contributions, while several others have been designed to reach ignition, but not built yet (IGNITOR, FIRE). Equivalent stellarator concepts, however, have barely been explored. The present study aims at filling this gap by: (1) performing an initial exploration of parameters relevant to ignition and of the difficulties for a high-field stellarator approach, and, (2) proposing a preliminary high-field stellarator concept for physics studies of burning plasmas and, possibly, ignition. To minimize costs, the device is pulsed, adopts resistive coils and has no blankets. Scaling laws are used to estimate the minimum field needed for ignition, fusion power and other plasma parameters. Analytical expressions and finite-element calculations are used to estimate approximate heat loads on the divertors, coil power consumption, and mechanical stresses as functions of the plasma volume, under wide-ranging parameters. Based on these studies, and on assumptions on the enhancement-factor of the energy confinement time and the achievable plasma beta, it is estimated that a stellarator of magnetic field B?~?10 T and 30 m³ plasma volume could approach or reach ignition, without encountering unsurmountable thermal or mechanical difficulties. The preliminary conceptual device is characterised by massive copper coils of variable cross-section, detachable periods, and a lithium wall and divertor.     

Systems studies of high-field tokamak ignition experiments

A study of the interaction between the physics of ignition and the engineering constraints in the design of compact, high-field tokamak ignition demonstration devices is presented. The studies investigate the effects the various electron and ion thermal diffusivities, which result from the many tokamak scaling laws, have on the design parameters of an ignition device and show the feasibility of building and igniting a compact tokamak (R < 1m). The relevant machine technology is discussed.     

Why Compact Tori for Fusion?

A compact torus (CT) has a toroidal magnetic and plasma geometry, but is contained within a simply-connected vacuum vessel such as a cylinder. Spheromaks and field-reversed configurations fall into this category. Compact tori are translatable and have a high engineering beta. The primary benefit of CTs for fusion is the absence of toroidal field and Ohmic Heating coils and the many problems brought on by them. Studying fusion-relevant plasma in simply-connected geometries affords the world fusion program both physics and technology opportunities not found in other configurations. This paper outlines the technology and physics opportunities of compact tori, and presents a cost model based on geometry for comparison with less compact configurations.     

Review of the magnetic fusion program by the 1986 ERAB Fusion Panel

The 1986 ERAB Fusion Panel finds that fusion energy continues to be an attractive energy source with great potential for the future, and that the magnetic fusion program continues to make substantial technical progress. In addition, fusion research advances plasma physics, a sophisticated and useful branch of applied science, as well as technologies important to industry and defense. These factors fully justify the substantial expenditures by the Department of Energy in fusion research and development (R&D). The Panel endorses the overall program direction, strategy, and plans, and recognizes the importance and timeliness of proceeding with a burning plasma experiment, such as the proposed Compact Ignition Tokamak (CIT) experiment.Presented to the Magnetic Fusion Advisory Committee (La Jolla, California, December 4, 1986)     

Report of the FESAC Panel on a Burning Plasma Program Strategy to Advance Fusion Energy

This panel was set up by the U.S. Department of Energy's Fusion Energy Sciences Advisory Committee in response to a request from the department to prepare a strategy for the study of burning fusion plasmas. Experimental study of a burning plasma has long been a goal of the U.S. science-based fusion energy program. There is an overwhelming consensus among fusion scientists that we are now ready scientifically, and have the full technical capability, to embark on this step. The fusion community is prepared to construct a facility that will allow us to produce this new plasma state in the laboratory, uncover the new physics associated with the fusion burn, and develop and test new technology essential for fusion power. Given this background, the panel has produced a strategy to enable the United States to proceed with this crucial next step in fusion energy science. The strategy was constructed with awareness that the burning plasma program is only one major component in a comprehensive development plan for fusion energy. A strong core science and technology program focused on fundamental understanding, confinement configuration optimization, and the development of plasma and fusion technologies essential to the realization of fusion energy. The core program will also be essential to the successful guidance and exploitation of the burning plasma program, providing the necessary knowledge base and scientific workforce.     

A Fusion Development Facility Based on the Spherical Torus

The Fusion community could soon have available an opportunity to significantly advance fusion development, an opportunity afforded by the potential of the spherical torus as a plasma confinement system. That opportunity is, in a single device at an affordable price and at a suitable site, to move sequentially through the major fusion objectives of advanced confinement physics, burning plasma and DT physics, blanket and other fusion nuclear technology development, tritium self-sufficiency, and, perhaps at the end with an upside performance outcome, a chance at net electric breakeven. Meeting this set of objectives would dramatically advance the development of fusion.     

Path to Market for Compact Modular Fusion Power Cores

The benefits of an energy source whose reactants are plentiful and whose products are benign is hard to measure, but at no time in history has this energy source been more needed. Nuclear fusion continues to promise to be this energy source. However, the path to market for fusion systems is still regularly a matter for long-term (20?+?year) plans. This white paper is intended to stimulate discussion of faster commercialization paths, distilling guidance from investors, utilities, and the wider energy research community (including from ARPA-E). There is great interest in a small modular fusion system that can be developed quickly and inexpensively. A simple model shows how compact modular fusion can produce a low cost development path by optimizing traditional systems that burn deuterium and tritium, operating not only at high magnetic field strength, but also by omitting some components that allow for the core to become more compact and easier to maintain. The dominant hurdles to the development of low cost, practical fusion systems are discussed, primarily in terms of the constraints placed on the cost of development stages in the private sector. The main finding presented here is that the bridge from DOE Office of Science to the energy market can come at the Proof of Principle development stage, providing the concept is sufficiently compact and inexpensive that its development allows for a normal technology commercialization path.     

A Novel superconducting toroidal field magnet concept using advanced materials

The plasma physics database indicates that two distinct approaches to tokamak design may lead to commercial fusion reactors: low Aspect ratio, high plasma current, relatively low magnetic field devices, and high Aspect ratio, high field devices. The former requires significant enhancements in plasma performance, while the latter depends primarily upon technology development. The key technology for the commercialization of the high-field approach is large, high magnetic field superconducting magnets. In this paper, the physics motivation for the high field approach and key superconducting magnet (SCM) development issues are reviewed. Improved SCM performance may be obtained from improved materials and/or improved engineering. Superconducting materials ranging from NbTi to high-T _c oxides are reviewed, demonstrating the broad range of potential superconducting materials. Structural material options are discussed, including cryogenic steel alloys and fiber-reinforced composite materials. Again, the breadth of options is highlighted. The potential for improved magnet engineering is quantified in terms of the Virial Theorem Limit, and two examples of approaches to highly optimized magnet configurations are discussed. The force-reduced concept, which is a finite application of the force-free solutions to Ampere's Law, appear promising for large SCMs but may be limited by the electromagnetics of a fusion plasma. The Solid Superconducting Cylinder (SSC) concept is proposed. This concept combines the unique properties of high-T _c superconductors within a low-T _c SCM to obtain (1) significant reductions in the structural material volume, (2) a decoupling of the tri-axial (compressive and tensile) stress state, and (3) a demountable TF magnet system. The advantages of this approach are quantified in terms of a 24 T commercial reactor TF magnet system. Significant reductions in the mechanical stress and the TF radial build are demonstrated.     

Future directions in fusion research: Super high-field tokamaks

Recent experimental results and advances in magnet engineering suggest that super high-field, high-aspect ratio tokamak devices could be a very efficient way to achieve burning plasma conditions and could open up a new area of research. Copper magnet devices with fields of 13 to 25 T at the plasma are considered. The super high-field approach could also provide advantages for ETR and demonstration/commercial reactor concepts (magnetic fields at the plasma in the 8–13 T range).Presented at the American Vacuum Society Meeting, October 30, 1986, Baltimore, Maryland.     

Plasma scenarios, equilibrium configurations and control in the design of FAST

The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed [A. Pizzuto on behalf of the Italian Association, The Fusion Advanced Studies Torus (FAST): a proposal for an ITER Satellite facility in support of the development of fusion energy, in: Proceedings of 22nd IAEA Fusion Energy Conference, Geneva, Switzerland, October 13–18, 2008; Nucl. Fusion, submitted for publication] as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in deuterium plasmas can be obtained from investigations of non linear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems.FAST equilibrium configurations have been designed in order to reproduce those of ITER with scaled plasma current, but still suitable to fulfil plasma conditions for studying burning plasma physics issues in an integrated framework. In this paper we report the plasma scenarios that can be studied on FAST, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. All plasma equilibria satisfy the following constraints: (a) minimum distance of 3 energy e-folding length (assumed to be 1 cm on the equatorial plane) between plasma and first wall to avoid interaction between plasma and main chamber; (b) maximum current density in the poloidal field coils, transiently, up to around 30 MA/m². The discharge duration is always limited by the heating of the toroidal field coils that are inertially cooled by helium gas at 30 K. The location of the poloidal field coils has been optimized in order to: minimize the magnetic energy; produce enough magnetic flux (up to 35 Wb stored) for the formation and sustainment of each scenario; produce a good field null at the plasma break-down (B_P/B_T < 2 × 10⁻⁴ at low field, i.e. B_T = 4 T and E_T = 2 V/m for at least 40 ms).Plasma position and shape control studies will also be presented. The optimization of the passive shell position slows the vertical stability growth time down to 100 ms.     

A programmatic framework for the Tokamak Physics Experiment (TPX)

Significant advances have been made in the confinement of reactor-grade plasmas, so that we are now preparing for experiments at the power breakeven level in the JET and TFTR experiments. In ITER we will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, we must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the U.S. domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX we can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. We can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current U.S. fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program.Abbreviations used TFTR Tokamak Fusion Test Reactor - JET Joint European Torus - ARIES Advanced Reactor Innovations Evaluation Study - SSTR Steady State Tokamak Reactor - PBXM Princeton Beta Experiment-Modified - DIII-D Doublet III—Dee - JT60-U Japanese Tokamak 60-Upgrade     

A Review of Confinement Requirements for Advanced Fuels

The energy confinement requirements for burning D-³He, D-D, or P-¹¹B are reviewed, with particular attention to the effects of helium ash accumulation. It is concluded that the DT cycle will lead to the more compact and economic fusion power reactor. The substantially less demanding requirements for ignition in DT (the n_{e E} T required for ignition in DT is smaller than that of the nearest advanced fuel, D-³He, by a factor of 50) will allow ignition, or significant fusion gain, in a smaller device; while the higher fusion power density (the fusion power density in DT is higher than that of D-³He by a factor of 100 at the same plasma pressure) allows for a more compact and economic device at fixed fusion power.     

Three Game Changing Discoveries: A Simpler Fusion Concept?

This paper encourages exploration of a broad range of magnetic fusion concepts in parallel with mainline tokamak development. Such exploration will certainly lead to increased understanding of fusion science and possibly to an attractive fusion energy concept. As an example, this paper describes three discoveries which greatly increase the attractiveness of the magnetic mirror plasma confinement concept. The mirror concept is thought to have three unattractive characteristics. The magnets are complex, the plasma is plagued with micro-instabilities and the electron temperature would never approach required keV levels. Persistent research on the gas dynamic trap device at the Budker Institute of Nuclear Physics in Russia and elsewhere have overcome these three deficiencies. Stable high energy density plasma can be confined with simple circular magnets, micro-instabilities can be tamed, and electron temperatures reaching a keV have been measured. These three accomplishments provide a basis to reconsider the mirror concept as a neutron source for medical applications, fusion materials development, nuclear fuel production, and fusion energy production.     

Could Advanced Fusion Fuels Be Used with Today's Technology?

Could today's technology suffice for engineering advanced-fuel, magnetic-fusion power plants, thus making fusion development primarily a physics problem? Such a path would almost certainly cost far less than the present D-T development program, which is driven by daunting engineering challenges as well as physics questions. Advanced fusion fuels, in contrast to D-T fuel, produce a smaller fraction of the fusion power as neutrons but have lower fusion reactivity, leading to a trade-off between engineering and physics. This paper examines the critical fusion engineering issues and related technologies with an eye to their application in tokamak and alternate-concept D-³He power plants. These issues include plasma power balance, magnets, surface heat flux, input power, fuel source, radiation damage, radioactive waste disposal, and nuclear proliferation.     

Nova target physics program and the Nova upgrade laser

Summary Recent advances in ICF target design and performance have made possible the achievement of ignition and gain with 1–2 MJ laser drive energy, as against the 5–10 MJ necessary to achieve high gain in the earlier designs. Ignition and propagating burn can be achieved at the lower energy by increasing the hohlraum temperature and, thereby increasing the pressure driving the imploding fusion capsule. Nova experiments continue to address the target physics of radiatively driven targets, such as laser-plasma interaction physics, the efficiency of laser light conversion to X-rays, hohlraum characterization and design, hydrodynamic stability, and implosion physics. Recent experiments on Nova have also demonstrated 1.3 times higher hohlraum temperature than previously predicted. This latter demonstration is the key achievement leading to the Nova Upgrade proposal. These combined results, together with those from experiments to study the interaction of high-power laser light with target plasmas, indicate that the capsule drive and symmetry conditions required for ignition and net gain can be achieved with a properly designed upgrade of the existing Nova facility.Success in the Nova Upgrade objective would firmly establish target and driver requirements for achieving high yield and high gain and would support a decision to construct a Laboratory Microfusion Facility (LMF) for defense applications and an Engineering Test Facility (ETF) for energy applications by the end of the first decade of the next century. Nova Upgrade experiments would focus on the target physics necessary to determine the minimum driver energy required to achieve ignition and high-gain laser fusion. The thermonuclear yield produced (up to 20 MJ) would be used to study the effects of fusion microexplosions on potential LMF and ETF reactor chamber materials. This information would permit development of the most efficient and least costly designs for the LMF and the ETF.In collaboration with W. H. Lowdermilk, N. Frank, C. D. Henning, John R. Murray, M. T. Tobin, J. R. Smith, E. K. Storm, J. D. Lindl, J. D. Kilkenny, J. T. Hunt, and J.B. Trenholme.     

Advantages of high-field tokamaks for fusion reactor development

High-field designs could reduce the cost and complexity of tokamak reactors. Moreover, the certainty of achieving required plasma performance could be increased. Strong Ohmic heating could eliminate or significantly decrease auxiliary heating power requirements and high values of n_E could be obtained in modest-size plasmas. Other potential advantages are reactor operation at modest values of , capability of higher power density and wall loading, and possibility of operation with advanced fuel mixtures. Present experimental results and basic scaling relations imply that the parameterB ²a, where B is the magnetic field and a is the minor radius, may be of special importance. A superhigh-field compact ignition experiment with very high values ofB ²a (e.g.,B ²a=150 T² m) has the potential of Ohmically heating to ignition. This short-pulse device would use inertially cooled copper plate magnets. Compact engineering test reactor and/or experimental hybrid reactor designs would use steady-state, water-cooled copper magnets and provide long-pulse operation. Design concepts are also described for demonstration/commercial reactors. These devices could use high-field superconducting magnets with 7–10 T at the plasma axis.     

Measuremeints on Current Sheets in Plasmas

The acceleration of a gas undergoing ionization in a crossed electric-magnetic field leads to the formation of current sheets similar to those observed in coaxial plasma guns. The dynamics of such sheets is of great interest to workers in plasma physics in view of their application to controlled thermonuclear fusion experiments and to space propulsion. In the present paper a detailed mapping of the magnetic field in the vicinity of a current sheet is described. Conventional small search coils are used for the detection of this field. From the configuration of this field an attempt is made to understand the processes taking place in the sheet. All of the experiments described in the paper are performed on a plasma coaxial gun of mall aspect ratio.     

Linear Connected Array of Non-Adiabatic Traps

An innovative confinement concept of a linearly connected array of non-adiabatic traps is examined. A non-adiabatic trap unit consists of a cylindrical vacuum chamber with external mirror coils and a pair of inner anti-parallel coils. Positions and currents on these coils are adjusted to achieve zero magnetic fields at the center. The plasma is trapped stably in this configuration due to the “Absolute Minimum B” magnetic configuration that is created. However, like a cusp field, energy confinement of plasma in a trap unit is poor, since particles suffer random pitch-angle scattering near the central zero field region and eventually fall into the loss-cone. However, once these non-adiabatic traps are linearly connected, plasma particles escaping a unit will be effectively re-trapped within the neighboring trap due to collision-free pitch-angle scattering in the zero field region. Since the transition of a charged plasma particle from unit to a next unit is stochastic, a connected array of non-adiabatic fusion core units and similar “leak suppressor” array units with low fusion rate at both ends of the core array improves the plasma confinement sufficient to achieve reactor-grade plasmas.     

Frontiers in Fusion Research

Presentations that were made at a Fusion Power Associates symposium, Frontiers in Fusion Research, are summarized. Discussions include the science frontiers of burning plasma physics, magnetic fusion concepts, inertial fusion concepts, and fusion technology.     

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation’s energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.