Recent advances in tokamak physics indicate the spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants. Full calculations of stability to kink and ballooning modes show the possibility of greater than 50% beta toroidal with the normalized beta [ p~= p~/ ( h b ) ] as high as 10 and fully aligned 100% bootstrap current. Such beta values coupled with 2-3 T toroidal fields imply a pilot plant about the size of the present DEI-D tokamak could produce -800 MW thermal, 160 M W net electric, and would have a ratio of gross electric power over recirculating power (&PLANT) of L9. The high beta values in the ST mean that ExB shear stabilization of turbulence should be 10 times more effective in the ST than in present tokamaks, implying that the required high quality of confinement needed to support such high beta values will be obtained. The anticipated beta values are so high that the allowable neutron flux at the blanket sets the device size, not the physics constraints. The ST has a favorable size scaling so that at 2-3 times the pilot plant size the QPLANT rises to 4-5, an economic range and 4 GW thermal power plants result. Current drive power requirements for 10% of the plasma current are consistent with the plant efficiencies quoted. The unshielded copper centerpost should have an adequate lifetime against nuclear transmutation induced resistance change and the low voltage, high current power supplies needed for the 12 turn TF coil appear reasonable. The favorable size scaling of the ST and the high beta mean that in large sizes, if the copper TF coil is replaced with a superconducting TF coil and a shield, the advanced fuel D-He3 could be burned in a device with Q p~m -4. If the anticipated physics of the ST regime can be proven in near term experiments and engineering challenges (such as the high power density to be exhausted and centerpost neutronics issues) can be met, then the ST offers the possibility of a magnetic fusion development path with a minimal cost initial step and exciting further possibilities.
The DIII–D National Fusion Facility at General Atomics (GA) focuses on plasma physics and fusion energy science. The DIII–D tokamak is a 35 m3 toroidal vacuum vessel with over 200 ports for diagnostic instrumentation, cryogenics, microwave heating, and four large neutral beam injectors. Maintaining vacuum in the 10−8 Torr range is crucial for producing high performance plasma discharges. He leak checking the DIII–D tokamak and the neutral beamlines has historically been difficult. D2 is used as the fuel gas in most plasma discharges and neutral beams. After plasma operations, D2 outgassing from the torus walls and internal beamline components can exceed 10−4 std cm3/s. The mass of the D2 molecule (4.028 u) is indistinguishable from that of the He atom (4.003 u) to a standard mass spectrometer leak detector. High levels of D2 reduce leak detector sensitivity and effectively mask the He trace gas signal rendering normal leak checking techniques ineffective. A simple apparatus was developed at GA to address these problems. It consists of a palladium based catalyst cell and associated valves and piping placed in series with the leak detector. This reduces the D2 throughput by a factor greater than 10 000, restoring leak detector sensitivity. This article will briefly discuss the development of the cell, the physical processes involved, the tests performed to quantify and optimize the processes, and the operational results at DIII–D.
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