Robert D. Woolley

National spherical torus experiment (NSTX) Center Stack Upgrade

The purpose of the NSTX Center Stack Upgrade project is to expand the NSTX operational space and thereby the physics basis for next-step ST facilities. The plasma aspect ratio (ratio of plasma major to minor radius) of the upgrade is increased to 1.5 from the original value of 1.26, which increases the cross sectional area of the center stack by a factor of ∼ 3 and makes possible higher levels of performance and pulse duration.

Tokamak poloidal magnetic field measurements accurate for unlimited time durations

A new hybrid method and apparatus is presented for poloidal magnetic field measurement, suitable for use in steady-state control of tokamak plasma shape, position, and current. [An invention disclosure has been filed.] It combines two different magnetic field principles (induction and torque) into a single hybrid measurement apparatus, thus in one device providing accurate magnetic field measurement over the entire frequency spectrum from DC to several kilohertz.

A novel demountable TF joint design for low aspect ratio spherical torus tokamaks

A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress.

Progress on a steady-state tokamak magnetic field sensor: accurate for unlimited time durations for fusion reactors

A novel hybrid method for accurate magnetic field measurements in steady-state or long-pulse tokamaks, described in a 1995 publication and patented in 1998, has not yet been tested in a tokamak. Recent development progress is described herein.

PF and TF power systems for the Fusion Ignition Research Experiment (FIRE)

The primary goal of the FIRE preconceptual design is to affordably conduct physics experiments in the fusion ignition regime with self-heated plasmas. The device could also permit advanced tokamak experiments in deuterium lasting several minutes. Tradeoff studies considering MVA and flattop duration have identified TF and PF magnet power system designs expected to have low cost, consistent both with the mission to enter the ignition regime and the possibility to conduct long pulse deuterium experiments. In addition, a performance-extending upgrade path for the power system has been identified which could be followed later, if experimental results justify further increasing the maximum toroidal field and plasma current or lengthening the duration.

Radial cooling of a Spherical Torus (ST) toroidal field (TF) centerpost

It is best for fusion to operate STs at the highest feasible toroidal field. This may not be obvious since dissipated electrical power increases as the square of magnetic field strength. However, fusion power density increases even faster. For example, increasing TF by 10% increases electrical losses by 21% but it also may increase fusion power by 46.4%. The main impediment blocking increased toroidal field is centerpost heat removal. Heat deposited by resistive heating and by radiation from the fusioning plasma is removed by coolant flowing through holes cut in the centerpost. Conventional designs have used axial flow within spaced, vertically oriented holes, and they have optimized flow speed, temperature rise, and cooling hole size. However, with all flow paths having the same length, the conductor volume removed is just the product of centerpost height by total flow area. Radial flow cooling is a radically different scheme promising a factor of almost two improvement over axial flow designs in the volume of conductor removed for cooling. Its performance advantage stems from its shorter average flow path length while retaining the same total flow cross sectional areas for inflow, outflow, and internal flows. This is accomplished by cooling the upper centerpost from the top and cooling the lower centerpost from the bottom, with no coolant crossing the horizontal midplane. For a single-turn TF, high pressure coolant is supplied both from the top and from the bottom to a central manifold located radially in the middle of the centerpost conductor. Coolant flows outward through many small diameter radially oriented cooling holes in the centerpost conductor into a low pressure annular manifold surrounding the centerpost. The external membrane surrounding the low pressure manifold includes sealed penetrations for the centerpost electrical connections and mechanical supports. Radial cooling optimization includes tapering of the manifold cross sections over their axial length in conjunction with varying the density and size of the radial cooling holes so that coolant flow speed is spatially constant and local cooling matches local heating. Radial cooling may simplify single-turn TF centerpost fabrication since it eliminates the need for long, narrow cooling holes as required for the axial schemes.

Flowing liquid lithium for the purpose of reducing tritium inventory levels in fusion energy reactors

A concern for fusion energy production reactors is tritium inventory resident in vacuum vessel enclosures during and after D-T operations. An immediate issue is the radiological safety associated with large quantities of tritium at risk. Additionally, there is an economic concern associated with the cost of tritium, having a current value in excess of

NSTX OH Coil Design Improvements

30 K/gram. Lithium has safely been deployed in fusion research reactors with good success. Our concept builds upon existing work, exploiting the ability of lithium to flow in toroidal and poloidal directions. In our deployment configuration, Li is used to bind with tritium deposited on main surfaces within the reactor vacuum vessel and ancillary internal components. Lithium can be used to wet appropriate surfaces within the vacuum vessel for the purpose of removing surface tritium from internal reactor structures, thus making it available for reuse as fuel.

Engineering Assessment of a National High-power Advanced Torus Experiment (NHTX)

The National Spherical Torus Experiment (NSTX) has been operating successfully since February of 1999. A unique element of NSTX is the center solenoid or OH coil that from the start has been a design challenged by the low aspect ratio/geometry of the device. To achieve this low aspect ratio the OH coils outer diameter is constrained to a narrow profile creating the need for creative design solutions concerning cooling connections, lead orientation, and insulation schemes. The original design has succeeded overall, but NSTX run time has been lost due to coil reliability issues. It was decided in the last year that it would be prudent to fabricate a new OH coil and have it available as an upgrade to the experiment. The experience of operating and maintaining the OH coil has provided the basis for an improved OH coil design. A collaboration was arranged with ASIPP in China to fabricate a spare coil for NSTX. The new OH coil will incorporate both design improvements intended to increase reliability as well as upgrades that will provide flexibility during future operation by allowing for an expanded operational profile. This paper summarizes and reviews these design and reliability improvements.

Operational Experience with NSTX Demountable TF Joint

A major challenge facing fusion development is the high heat flux power handling of plasma exhaust, for which a defining parameter is P/R, the ratio of exhaust power P to the radius R at which the divertor is located. Preliminary studies indicate that a compact, cost effective device can be constructed at PPPL exploiting existing infrastructure which could operate at P/R ~50, well in excess of levels available elsewhere and including future ITER operations. The mission for a National High-power Advanced Torus Experiment (NHTX) would be to study the integration of a high P/R plasma-boundary interface with high-confinement, high-beta, non-inductive plasma operation.