Cambridge MA (SPX) Dec 07, 2004

MIT and Columbia University students and researchers have begun operation of a novel experiment that confines high-temperature ionized gas, called plasma, using the strong magnetic fields from a half-ton superconducting ring inside a huge vessel reminiscent of a spaceship.

The experiment, the first of its kind, will test whether nature's way of confining high-temperature gas might lead to a new source of energy for the world.

First results from the Levitated Dipole Experiment (LDX) were presented at a meeting of the American Physical Society the week of Nov. 15. Scientists and students described more than 100 plasma discharges created within the new device, each lasting from 5 to 10 seconds.

X-ray spectroscopy and visible photography recorded spectacular images of the hot, confined plasma and of the dynamics of matter confined by strong magnetic force fields.

A dedication for LDX, the United States' newest approach to nuclear fusion, was held in late October. Fusion energy is advantageous because its hydrogen fuel is practically limitless and the resulting energy would be clean and would not contribute to global warming as does the burning of fossil fuels. Scientists using the LDX experiment will conduct basic studies of confined high-temperature matter and investigate whether the plasma may someday be used to produce fusion energy on Earth. Fusion energy is the energy source of the sun and stars. At high temperature and pressure, light elements like hydrogen are fused together to make heavier elements, such as helium, in a process that releases large amounts of energy.

Introduction

The dipole magnetic field is the simplest and most common magnetic field configuration in the universe. It is the magnetic far-field of a single, circular current loop, and it represents the dominate structure of the middle magnetospheres of magnetized planets and neutron stars. The use of a dipole magnetic field generated by a levitated ring to confine a hotp lasma for fusion power generation was first considered by Akira Hasegawa after participating in the Voyager 2 encounter with Uranus [1]. Hasegawa recognized that the inward difusion and adiabatic heating that accompanied strong magnetic and electric fluctuations in planetary magnetospheres represented a fundamental property of strongly magnetized plasmas not yet observed in laboratory fusion experiments. For example, it is well-known that global fluctuations excited in laboratory fusion plasmas result in rapid plasma and energy loss. In contrast, large-scale fluctuations induced by sudden compressions of the geomagnetic cavity (due to enhancements in solar wind pressure) or by unsteady convections occurring during magnetic substorms energize and populate the energetic electrons trapped in the Earth�fs magnetosphere [2]. The fluctuations induce inward particle diffusion from the magnetospheric boundary even when the central plasma density greatly exceeds the density at the edge. Hasegawa postulated that if a hot plasma having pressure profiles similar to those observed in nature could be confined by a laboratory dipole magnetic field, this plasma might also be immune to anomalous (outward) transport of plasma energy and particles.

The dipole confinementc oncept is based on the idea of generating pressure profiles near marginal stability for low-frequency magnetic and electrostatic fluctuations. For ideal MHD, marginal stability results when the pressure profile, p satis.es the adiabaticity condition, d(pV ƒÁ) = 0, where V is the flux tube volume (V =
d/B) and . = 5/3.

This condition leads to dipole pressure profiles that scale with radius as r-20/3, similar to energetic particle pressure profiles observed in the Earth�fs magnetosphere. Since the magnetic field of a dipole is poloidal, there is no drift off of flux surfaces and therefore no "neo-classical" degradation of confinement as seen in a tokamak. It has been pointed-out that a plasma that satisfies the MHD interchange stability requirement may be intrinsically stable to drift frequency modes. Stability of low frequency modes can be evaluated using kinetic theory and a Nyquist analysis permits an evaluation of stability boundaries with a minimum of simplifying assumptions. Using kinetic theory we have shown that when the interchange stability requirement (for small Larmor radius) becomes .*p > .p with .*p the diamagnetic drift frequency and .d the curvature drift frequency and this result is consistent with MHD[3]. This property implies that the pressure scale length exceed the radius of curvature, which is a physical property that distinguishes a dipole confined plasma from other approaches to magnetic fusion plasmas. Additionally, when the interchange stability criterion is satisfied, it can be shown that localized collisionless trapped particle modes and dissipative trapped ion modes become stable. Low frequency modes that are driven by parallel dynamics (i.e. the universal instability) also tends to be stable due to the requirement that the parallel wavelength of the mode fit on the closed field lines. Recent theoretical work on anomalous inward diffusion (towards the ring) due to high frequency, drift-cyclotron instability supports the view that both stability and confinement can be extremely good in a levitated dipole [4].

By levitating the dipole magnet end losses can be eliminated and conceptual reactor studies supported the possibility of a dipole based fusion [5, 6] power source that utilizes advanced fuels. The ignition of an advanced fuel burning fusion reactor requires high beta and good energy confinement. Additionally advanced fuels require steady state and efficient ash removal. A levitated dipole may provide uniquely good properties in all of these areas. The chief drawback of the dipole approach is the need for a levitated superconducting ring internal to the plasma and this provides a challenge to the engineering of the device. [hmmm - sounds like the isolated ring would coock in short order, in the 100-million-degree environment....] A fusion reactor based on a levitated dipole has been explored in two studies [5, 7]. Recent advances in high temperature superconductors coupled with an innovative design concept of Dawson [8] on the maintenance of an internal superconducting ring in the vicinity of a fusion plasma lead us to believe that this issue is technologically solvable.

The dipole confinement approach can be tested in a relatively modest experiment which profits form the development of the technology of superconductors, gyrotrons and pellet injectors. A concept exploration experiment is presently being developed jointly by Columbia University and MIT.

Teller and co-workers [7] developed a conceptual design of a levitated dipole space propulsion system.

Much of the heat that is incident onto the surface of the ring (either from particles or radiation) is expected to be radiated from the ring surface to the vacuum chamber wall. There is some heat leak into the superconductor and in a laboratory experiment the superconductor can be levitated for several hours before re cooling. 

In a reactor environment an inertially cooled ring could float for tens of hours. It has also been proposed that the ring in a reactor could contain internal refrigerators [8] and as a result the operation could be steady state. The development of high temperature superconductors would substantially ease the dificulty of designing the internal ring. A high TC (high critical temperature) coil would have an increased heat capacity and could be maintained for a relatively long pulse without the need for internal refrigerators. Furthermore a refrigerated steady state coil would be substantially easier to design using a high TC superconductor.

The dipole reactor concept is a radical departure from the better known toroidal-based magnetic fusion reactor concepts. For example, the most difficult problems for a tokamak reactor are the divertor heat dissipation, disruptions, steady state operation, and an inherently low beta limit. Furthermore, the tokamak is subject to neoclassical effects and, in many cases, anomalous, fluctuation-induced transport. The dipole concept provides a approach to fusion which solves these problems.

- Divertor problem: The diffculty in spreading the heat load at the divertor plate is generic to concepts in which the magnetic flux is trapped within the (toroidal field) coil system. By having the plasma outside of the confining coil the plasma flux can be sufficiently expanded to substantially reduce divertor and first wall heat loads.

- Major disruptions: A tokamak has a large amount of energy stored in the plasma current. The dipole plasma carries only diamagnetic current and is inherently free of disruptions. Furthermore there is evidence that when the dipole becomes MHD unstable, i.e. .p > .pcrit, the plasma will expand sufficiently to reduce the pressure gradient (much like tokamak type I ELMs). Therefore MHD instability will not lead to a loss of plasma.

- Steady state: A tokamak is a pulsed device and current drive schemes that are required for steady operation appear to be costly. The dipole plasma is inherently steady state.

- Beta limits: Tokamak stability depends on the poloidal field which is less than the toroidal .eld by Bp/BT ~ a/qR and ßpol ~ 1 determines a beta limit ß  1. For a dipole there is a critical pressure gradient that can be supported and for a sufficiently gentle pressure gradient the dipole plasma resides in an absolute energy well and is stable up to local beta values in excess of unity.

- Transport and neoclassical effects: The trapping of particles in regions of bad curvature makes the tokamak susceptible to drift frequency range turbulent transport. A dipole can, theoretically, be stable to low frequency drift modes [1, 3]. In addition a tokamak has a "neoclassical" degradation of transport that derives from the drifts of particles off of the flux surfaces. In a dipole the drifts are toroidal and they define the flux surfaces. Therefore the irreducible minimum transport for a dipole is governed by the "classical" and not the "neoclassical" limit.

- Fueling: Fueling and ash removal are an important issue in an ignited reactor and are of particular importance for operation with advanced fuels which deposit all of the fusion products within the fusing plasma. Magnetic confinement configurations that do not have shear may be subject to convective cells [9, 18]. At the critical pressure gradient for marginal stability the resulting convective flows can transport particles without a net transport of energy, i.e. the hot core plasma cools as it convects outwards and the outer plasma heats as it convects inward. This would provide the ideal approach for fueling a reacting fusion plasma. The dipole reactor would also provide signifficant and signifficantly different engineering challenges from a tokamak. The outstanding issues are the sustenance of a superconducting coil embedded within a fusing plasma. This issue plus the advantages listed above such as high beta, good confinement and improved fueling makes the dipole concept particularly well suited for advanced fuels. The use of warm (high TC) superconductors and the development of creative ideas for internal refrigerators [8] would greatly enhance the feasibility of this approach.

Much work remains to be done to evaluate the compatibility of a dipole configuration with advanced fuel reactors. One fuel cycle that is desirable because of the abundance of fuel is the DD cycle. An other interesting fuel cycle would utilize D and 3He. However, since 3He does not occur naturally in sufficient quantity on the earth, the utilization of the D3He cycle (on the earth) would require the mining and transportation of 3He from the moon. [ooops.... here we go again !]

J. Kesner, D.T. Garnier�õ, A. Hansen�õ, M. Mauel�õ, L. Bromberg

The D-T rate coefficient is two orders of magnitude larger than the rate coefficient for Deuterium-3Helium (D-3He) reaction or for the deuterium-deuterium (D-D) reaction.

The D-D reaction is perhaps the most interesting from the point of view of eliminating both the tritium and the energetic neutron problems. However the relatively small fusion cross section has made this approach problematical.  A direct consequence of the low reactivity is that the buildup of ash in the fusing plasma can preclude ignition in a tokamak-like device [1].

In this study we show that a levitated dipole device may be ideally suited for a D-D based fusion power source.

The most important reactions for controlled nuclear fusion are as follows:

D + T ---> 4He(3.5 MeV ) + n(14.1 MeV )

D +3 He ---> 4He(3.6 MeV ) + p(14.7 MeV )

D + D 50% ---> 3He(0.82 MeV ) + n(2.45 MeV )

D + D 50% ---> T(1.01 MeV ) + p(3.02 MeV )

Referring back to the fusion reactions shown in Eq. (1) there are two equally likely D-D fusion reactions. The first reaction produces a 3He whereas the second produces a triton. The 3He will fuse with the background deuterium. Permitting the tritium to fuse leads to the "catalyzed DD" fuel cycle. However because the D-T reaction would produce an energetic (14.1MeV) neutron that would be difficult to prevent from entering and heating an internal coil, we propose to remove the triton before a substantial fraction can fuse and replace it with the 3He tritium decay product. This leads to the production of 22 MeV of energy per D-D fusion reaction. This fusion cycle has been discussed in References [11, 12] and will be referred to as "Helium catalyzed D-D" fusion.

The energetic triton produced in the primary D-D reaction, however, will slow down and thermalize before it can be convected out to the plasma edge and removed. We can estimate the beam-plasma fusion rate for an energetic triton slowing down in a thermal deuterium plasma. Using the energy loss rate from Ref. [16] and the D-T fusion cross-sections from Ref. [17] we can obtain the fusion probability for a 1 MeV triton in a warm deuterium plasma. The fusion probability as a function of plasma temperature (Te = Ti) is shown in Fig. 2. For a 40 KeV deuterium plasma we find that approximately 7% of the tritons fuse as they slow down.

The surface power loading is relatively low (< 0.1 MW/m2) and suggestive of a relatively thin-wall vessel containing a slowly flowing coolant. The internal floating coil will operate with a high outer surface temperature (> 16000K) so as to radiate away all of the surface and neutron heating via black body radiation (assuming an emissivity ~1). In addition, we envision that the floating coil will have internal refrigerators that will pump to the surface the heat that is deposited directly into the superconducting coil via volumetric neutron heating.

Vac. vessel midplane radius (m) 30

First wall volume (m3) �õ 269

Fusion Power, Pfus (MW) 610

Bremsstrahlung radiation (MW) 430

The floating coil consists of a winding pack surrounded by a cryostat that provides both thermal and neutron shielding. For steady state operation the floating coil must include an internal refrigerator in order to maintain the superconductor at a low temperature. For such a design it is critical to minimize the power deposited into the superconducting coil from volumetric neutron heating as it is inefficient to extract heat from the cold coil (~70 0K)and deposit it on the hot outer surface of the coil

<SNIP>

Equation (11) indicates that the outer surface of the coil will rise to an average temperature, Tsurf ~1, 800 0K. The low thermal efficiency associated with maintaining the superconductor

at a low temperature will require that a great deal of attention be focused on the design of the floating coil shield. ........The best results (least direct heating of superconductor) were found for the latter segmented shield which indicates a direct deposition into the coil of 1.4 KW from high energy and 2.2 KW from low energy neutrons. The low level of heating from the 14 MeV neutrons requires the removal of thermal tritium as we have assumed. In total we find that there is 137 MW of power deposited into the surface of the coil (DD study in Table 4)...........Maintaining a superconducting ring within a fusing plasma is a challenging task. One must design of refrigerator that can eject heat at above 1600 0K.  Furthermore the refrigerator must be powered by a generator that operates between the high temperatures of the outer shell of the floating coil, i.e. between 1500-1600 0K and 1800-1900 0K .

The storage of the of the tritium that is removed from the discharge during its 12.3 year half life will require the safe storage of 100 to 200 Kg of tritium.