The Plasma Fusion Center is recognized as the leading university laboratory in developing the scientific and engineering aspects of magnetic confinement fusion and related plasma science and technology. Its research programs continue to produce significant results on several fronts: (a) experimental confinement research on the Alcator C-Mod tokamak (investigations of the stability, heating, and transport properties of compact high magnetic field, diverted plasmas), (b) the basic physics of energetic plasmas (plasma theory, theoretical support of ITER, TPX and IGNITOR, new confinement concepts, nonneutral plasmas, coherent EM wave generation, development of high-temperature plasma diagnostics, basic laboratory and ionospheric plasma physics experiments, and novel diagnostic of inertial fusion experiments), (c) a broad program of fusion technology and engineering development that addresses problems in several areas (e.g., magnetic systems, superconducting materials, fusion environmental and safety studies, advanced millimeter-wave sources, system studies of fusion reactors, including operational and technological requirements), and (d) the growing activity of plasma waste treatment using plasmas.
Approximately 40 percent of the Center's activities are associated with the Alcator C-Mod tokamak experiment, 40 percent with the research on superconducting magnet system components for future fusion devices, and the remaining 20 percent with the many other R& D activities.
The Plasma Fusion Center R&D programs are supported principally by the Department of Energy's Office of Fusion Energy. There are approximately 305 personnel associated with PFC research activities. These include: 23 faculty and senior academic staff, 61 graduate students and 20 undergraduate students, with participating faculty and students from Electrical Engineering and Computer Science, Materials Science and Engineering, Mechanical Engineering, Nuclear Engineering, and Physics; 96 research scientists and engineers and 40 visiting scientists; 33 technical support personnel; and 32 administrative and support staff.
The Plasma Fusion Center's high-field tokamak, Alcator C-Mod, started full operation in May 1993. It was designed with the capability to address a range of issues, foremost including ITER-relevant research topics, as well as physics issues affecting the design of an advanced (long pulse) tokamak fusion reactor. A particularly difficult problem for ITER is the design of a divertor, or heat removal system, which will exhaust several hundred megawatts of thermal power. Alcator C-Mod has an advanced divertor design, unique among presently operating tokamak experiments, incorporating many of the key elements in the present ITER design approach. Thus, the critical issues of particle and power exhaust are a central program focus.
PFC programs also support ITER in critical technology areas, including superconducting magnets and development of millimeter wave RF sources suitable for heating and driving current near the electron gyrofrequency. In the magnetics area, the PFC leads the US ITER-Home Team magnetics effort in an extensive, internationally coordinated program of superconducting magnet development leading to construction of magnets on a scale and performance level well beyond that of present-day experience.
Earlier this year Prof. Miklos Porkolab was chosen to replace Prof. Ron Parker as Director of the PFC. Dr. Bruce Montgomery will remain Associate Director for Engineering and Technology, while the Director for Research position remains to be filled. Mr. Willie Smith (previously of the Provost's Office) has been appointed as the new Administrative Officer.
Alcator C-Mod is now established as one of the "big five" divertor tokamaks in the world. Two of the others, JET and JT-60U are the large "flagships" of their national programs, whereas ASDEX-U and DIII-D are medium sized, lower field devices. Alcator C-Mod is the only high-field compact experiment of the group and therefore plays a unique role in providing critical tests of scaling and fusion theory at high power density. Alcator C-Mod is extremely well placed to contribute to solving problems of ITER and future fusion reactors. Its vertical plate divertor geometry has been adopted as ITER's reference design, and its shape is essentially the same as ITER's. It also is presently unique in having high-Z metallic plasma facing components. Such components are planned for ITER.
Alcator C-Mod is thus extremely favorably placed to contribute vital information to fusion research, and to do so in a highly cost effective way because of its compact, high field approach. Nevertheless, serious concern exists about the continued funding of the project because of the large cuts in the fusion program presently being proposed in Congress. We believe that even in a restructured fusion program, Alcator C-Mod has a uniquely important role to play. Alcator C-Mod, with its total staff in excess of 100 people, including 18 full-time research physicists, 2 faculty members, and 20 graduate students, makes a unique and essential contribution to the MIT educational program. Naturally, PFC staff are sparing no effort to point out the opportunities that Alcator C-Mod represents and to argue for its continuation and expansion.
A systematic technique for controlling the plasma shape using orthogonal controllers has been developed and successfully implemented. This technique has improved the robustness of our plasma control system, and facilitated development of new equilibria, including limited, as well as lower and upper single null divertor geometries; elongations have been varied from 0.9 to 1.85.
In a series of experiments with potential significance for future fusion facilities, remote operation of a tokamak was demonstrated for the first time. A team of M.I.T. scientists, in collaboration with a team from LLNL, operated Alcator C-Mod from a remote control room in Livermore, California, with control and shot data as well as interpersonal communications being transmitted over the Internet.
Studies of disruptions have continued, with special attention to halo currents, which flow partly in the plasma and partly through the conducting vessel. The ratio of halo current to plasma current has been shown to be proportional to 1/q. Toroidal asymmetries and rotation of the halo currents have been measured, and a correlation between halo currents and the occurrence of integral values of an effective rotational transform, including the current path through the wall, has been observed.
Transport experiments have been carried out with ICRF heating up to 3.5 MW. The L-mode plasmas have confinement close to the empirical ITER-89P scaling. H-mode operation, with improved particle and energy confinement, has been achieved at fields up to 8 Tesla, with a range of plasma and surface power density spanning those characteristic of ITER or a reactor. The threshold power density for transition to H-mode is found to scale with density and field, similar to observations on other experiments. However, in C-Mod the observed threshold values are up to a factor or two lower than those obtained elsewhere, while the surface power density in C-Mod exceeds that in other machines by a factor of ten.
Experiments have begun to test the principles of dimensionless scaling of tokamak transport by running discharges with identical geometry, collisionality, and gyro-size in C-Mod and DIII-D, a larger, lower-field tokamak at General Atomics. These experiments should help validate projections to future devices by extrapolations in gyro-size from current facilities.
In the last 12 months excellent progress has been made at Alcator C-Mod both in terms of experimental capabilities and in divertor physics R&D for ITER. In particular, the present divertor diagnostic complement delivers more detailed information about divertor characteristics (radiation, density, temperature) than is available from any other tokamak. Tomographic inversion techniques have been applied to divertor bolometry and visible imaging to determine the strength and spatial distribution of total radiated power and of particular species in the divertor region. We routinely have complete between-shot Langmuir probe density and electron temperature profiles at the divertor plate and in the tokamak edge plasma SOL (scrape-off layer). Important studies of divertor detachment have shown that the detachment threshold can be modified through the controlled addition of impurity gases (methane and neon) as well as by changing the divertor geometry. The flat-plate divertor geometry has a significantly higher detachment threshold (factor of ~ 2) compared to the vertical plate or slot geometry. This result, which is consistent with the finding that Alcator C-Mod typically has a lower detachment threshold (compared to the "Greenwald density" limit) than other tokamaks, all of which had operated with flat-plate divertors, has important implications for the design of the ITER divertor.
The diagnostics for Alcator C-Mod have reached a mature level of development, with accurate measurements of the majority of important plasma parameters available on a routine basis. In the area of advanced diagnostic development, we are concentrating on upgrading our divertor diagnostic capabilities, along with providing enhancements to interferometry and fluctuation measurements. Much of the diagnostic development proceeds through collaboration with outside groups, including Princeton Plasma Physics Lab (spectroscopy, divertor Thomson scattering, interferometry, reflectometry), Oak Ridge National Lab (CO2 scattering, faraday polarimetry), Johns Hopkins University (spectroscopy, atomic physics), and the University of Maryland (spectroscopy).
PHYSICS RESEARCH DIVISION
Reports to the President 1994-95
MIT Reports to the President 1994-95