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The current drive experiment-upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego (UCSD). Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance.

Results on the interaction between plasma in the current drive experiment-upgrade (CDX-U) spherical torus and a liquid lithium limiter are reported. It is observed that macroscopic lithium droplets detach from the limiter head and fall towards the plasma core. However, no disruptions occurred during these discharges despite the fact that relatively large-scale blobs are observed entering the confined plasma. A multi-tip Langmuir probe measures the edge plasma properties. It is found that the average density and temperature and their fluctuations are unaffected by the presence of lithium within experimental error.

A flowing liquid lithium first wall or divertor target could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls in fusion reactors. To investigate the interaction of a spherical torus plasma with liquid lithium limiters, large area divertor targets, and walls, discharges will be established in the Current Drive Experiment-Upgrade (CDX-U) where the plasma–wall interactions are dominated by liquid lithium surfaces. Among the unique CDX-U lithium diagnostics is a multilayer mirror (MLM) array, which will monitor the 13.5 nm LiIII line for core lithium concentrations. Additional spectroscopic diagnostics include a grazing incidence extreme ultraviolet (XUV) spectrometer (STRS) and a filterscope system to monitor Dα and various impurity lines local to the lithium limiter. Profile data will be obtained with a multichannel tangential bolometer and a multipoint Thomson scattering system configured to give enhanced edge resolution. Coupons on...

The experimental program of the Princeton Beta Experiment‐Modified (PBX‐M) tokamak [Phys. Fluids B2, 1271 (1990)] is directed toward tailoring plasma profiles to achieve greater stability and confinement and to gain access to the second stability region. Modification of the current density profile has been achieved with lower‐hybrid current drive (LHCD), leading to a regime free of global magnetohydrodynamic modes, while raising the value of q(0) above unity. The diffusion of the fast electrons produced by LHCD has been examined using two‐dimensional hard x‐ray imaging. Ion Bernstein waves (IBW) have been used for ion heating: a preliminary analysis shows that ion heating was spatially localized and in agreement with theoretical calculations. Divertor biasing has modified the electric field inside the last closed surface, resulting in the formation of a transport barrier, which in turn has reduced the threshold power of neutral beam injection (NBI) for H‐mode transition by 25%.

Recent experiments in the Current Drive Experiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, B{sub toroidal} = 2 kG, I{sub P} = 100 kA, T{sub e}(0) = 100 eV, n{sub e}(0) {approx} 5 x 10{sup 19} m{sup -3}) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm{sup 2} (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Z{sub effective} < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiationmore » are strongly reduced.« less

Two conceptual designs were investigated for an experimental facility with sufficient versatility to evaluate active mode stabilization of high power tokamak edge plasmas. This paper discusses the experimental required to innovative methods for edge control stabilization.

Summary form only given. The new Helimac plasma device will produce a plasma that lends itself to the study of plasma turbulence. The configuration is accessible to straightforward theory of drift wave turbulence. The device is toroidal and has vertical and toroidal magnetic fields. Though it shares some features with a tokamak, it is substantially colder, lower density, and has a simpler magnetic configuration than do typical tokamaks. These properties make it easier to diagnose. The Helimac approximates an infinite cylinder with an MHD equilibrium that depends on a single radial variable. The magnetic field lines are helices of tight pitch. The configuration is intrinsically one dimensional with both magnetic curvature and shear. Flow shear can be externally applied and controlled. With an average radius of 1 m and height of 2 m, the size is very large compared with all scale lengths at the field of 0.1 T, temperatures of 10 eV, and densities of 10/sup 17/ m/sup -3/. The device to be built will feature continuous operation. The diagnostics will consist of full electrostatic probe diagnostics for complete characterization of the turbulence and of spectroscopic measurements of the induced flow field. The experimental design and plans for turbulence measurement and flow-shear suppression will be presented along with the conceptual design for the major diagnostics.

The Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego. Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance.

We discuss the fluctuation characteristics expected from MHD modes including ballooning modes and show how they can be studied using measurements of ECE. A novel technique, ECE correlation radiometry, which enables the location and spatial structure of long wavelength (L≳0.1×a) MHD modes to be determined, is described. Measurements on JET with a 44‐channel ECE heterodyne radiometer are presented and show the existence of high‐frequency (high n number) MHD modes under high poloidal beta, pellet‐enhanced performance (PEP) plasma conditions. Similar measurements are planned for PBX‐M and the measurement system is described.

A flowing liquid lithium first wall or diverter target could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls in fusion reactors. To investigate the interaction of a spherical torus plasma with liquid lithium limiters, large area diverter targets, and walls, discharges will be established in the Current Drive Experiment-Upgrade (CDX-U) where the plasma-wall interactions are dominated by liquid lithium surfaces. Among the unique CDX-U lithium diagnostics is a multi-layer mirror (MLM) array, which will monitor the 135 {angstrom} LiIII line for core lithium concentrations. Additional spectroscopic diagnostics include a grazing incidence XUV spectrometer (STRS) and a filterscope system to monitor D{sub {alpha}} and various impurity lines local to the lithium limiter. Profile data will be obtained with a multichannel tangential bolometer and a multipoint Thomson scattering system configured to give enhanced edge resolution. Coupons on th e inner wall of the CDX-U vacuum vessel will be used for surface analysis. A 10,000 frame per second fast visible camera and an IR camera will also be available.

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new groundbreaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only approximately1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment.

A third‐harmonic electron cyclotron emission diagnostic using ultrawide bandwidth (≊40 GHz) heterodyne receivers centered on 120 GHz with 14 channels per radial view is described for localized, long wavelength (5≲λ≲50 cm), fast time response (≊1 μs) fluctuation studies in the PBX‐M tokamak. The optically gray emission signal will have a γne/ne+(3/β)Te/Te dependence on temperature and density fluctuations where γ≤1 and 1≤β≤3 depending on local optical depth. Electron temperature fluctuation sensitivity is estimated to be 0.2%≤Te/Te ≤2.9% depending on local optical depth and fluctuation frequency in the 0.1–1 MHz range. Spatial resolution of approximately 3 cm radially and 5 cm vertically are estimated for 2 keV plasmas with low suprathermal electron emission.

The TST is a small steady-state tokamak designed for testing divertors under conditions similar to those anticipated in future large tokamaks. An initial design has R0/a = 2.5, R0 = 0.75 m, a = 0.3 m, and Bt0 = 2.2 T with full inductive capability. With heating and current drive power of 4.5 MW, the heat flux at the plasma edge Q⊥ can be as high as 0.3 MW/m2. Plasma currents Ip above 500 kA can be maintained by 1 MW of lower hybrid power (2.45 GHz) for average densities ne up to 3 × 1019 m−3. Additional power via ICRF (2 MW) and neutral beams (1.5 MW) is needed to maintain current for ne up to 5 × 1019 m−3. Fully demountable, actively cooled, steady-state toroidal field coils permit ample access for the auxiliary systems and divertor cassettes. The toroidal field magnets require a steady-state supply of less than 40 MW. The size and cost of the TST can be reduced by eliminating the solenoid, reducing Bt0 to 1.4 T, and lowering R0/a to 1.7. This option permits low-R0/a experimentation while maintaining the capability for testing divertors but requires successful noninductive current initiation and maintenance in the low-R0/a regime.

Localized measurements of internal magnetohydrodynamic (MHD) activity in the Princeton Beta Experiment-Modification (PBX-M) have been made using a heterodyne radiometer. The radiometer measures third harmonic optically gray electron cyclotron emission (ECE) from the center to the outer edge of the plasma on eight radially localized frequency channels from 112–128 GHz. The radiometer uses a focusing Gaussian optics system, which provides a vertical resolution of approximately 6 cm throughout the plasma. The radial resolution of the diagnostics is limited by the relativistic broadening to approximately 2–3 cm. Long wavelength (k