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Various perfluorinated materials are used at the Princeton Plasma Physics Laboratory in support of the Deuterium-Tritium experimental program on the Tokamak Fusion Test Reactor (TFTR). For example, SF{sub 6} is used as a high dielectric gaseous insulator in the Neutral Beam sources, and Krytox, a perfluorinated polyether, is used as a lubricant in vacuum pumping systems. Each of these materials is robust and stable in the applications for which they are designed but may be a source of trouble when used in tritium systems. This paper reports on the observations made and experience gained operating tritium systems under conditions which degrade these perfluorinated materials. The possible degradation mechanisms and products are described, and the effect on the equipment and instrumentation is described. These observations have led to the conclusion that under certain circumstances perfluorinated materials are not suited for tritium service because of the degradation products from tritium decay and/or process conditions. 11 refs., 1 fig., 1 tab.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fullymore » explored.« less

A conceptual design for a magnetic intervention system is presented in support of a 2 GW IFE direct drive fusion power reactor. The system is designed employing a cusp field to deflect ions generated by an IFE implosion away from the first wall of the reactor core and into specifically designed ion dumps. The magnetic coil system will employ liquid helium cooled 5083 Aluminum alloy casing on a Rutherford NbTi cable. The cables are configured as four double pancakes with a 5083 Aluminum alloy case for structural support. The conceptual design and corresponding preliminary load and field calculations will be presented.

The lithium pellet injector (LPI) is being developed for the National Spherical Torus Experiment (NSTX). The LPI will inject 'pellets' of various composition into the plasma in order to study wall conditioning, edge impurity transport, liquid limiter simulations, and other areas of research. The control system for the NSTX LPI has incorporated widely used advanced technologies, such as LabVIEW/spl trade/ and PCI bus I/O boards, to create a low-cost control system which is fully integrated into the NSTX computing environment. This paper will present the hardware and software design of the computer control system for the LPI.

A supersonic gas injector (SGI) has been developed for fueling and diagnostic applications on the National Spherical Torus Experiment (NSTX). It is comprised of a graphite converging-diverging Laval nozzle and a commercial piezoelectric gas valve mounted on a movable probe at a low field side midplane port location. Also mounted on the probe is a diagnostic package: a Langmuir probe, two thermocouples and five pick-up coils for measuring toroidal, radial, vertical magnetic field components and magnetic fluctuations at the location of the SGI tip. The SGI flow rate is up to 4times1021 particles/s, comparable to conventional NSTX gas injectors. The nozzle operates in a pulsed regime at room temperature and a reservoir gas pressure up to 0.33 MPa. The deuterium jet Mach number of about 4, and the divergence half-angle of 5deg - 25deg have been measured in laboratory experiments simulating NSTX environment. In initial NSTX experiments reliable operation of the SGI and all mounted diagnostics at distances 1-20 cm from the plasma separatrix has been demonstrated. The SGI has been used for fueling of ohmic and 2-4 MW NBI heated L- and H-mode plasmas. Fueling efficiency in the range 0.1-0.3 has been obtained from the plasma electron inventory analysis

A pellet injector suitable for the injection of lithium and other low-Z pellets of varying mass into plasmas at precise velocities from 5 to 500 m/s is being developed for use on NSTX. The ability to inject low-Z impurities will significantly expand NSTX experimental capability for a broad range of diagnostic and operational applications. The architecture employs a pellet-carrying cartridge propelled through a guide tube by deuterium gas. Abrupt deceleration of the cartridge at the end of the guide tube results in the pellet continuing along its intended path, thereby giving controlled reproducible velocities for a variety of pellets materials and a reduced gas load to the torus. The planned injector assembly has four hundred guide tubes contained in a rotating magazine with eight tubes provided for injection into plasmas. A PC-based control system is being developed as well and will be described elsewhere in these Proceedings. The development path and mechanical performance of the injector will be described.

The Tokamak Fusion Test Reactor (TFTR) has twenty toroidal-field (TF) coils with a maximum current rating of 73 kA for 3.6 s when cooled with water. Operation of the TF array at this maximal current produces a 5.2-T magnetic field at a major radius of 2.48 m. Repetition rate is dependent upon the efficiency of the cooling system for the array. An alternative approach to the water cooling would be to circulate a silicate ester (oil) in the coils. This technique would improve the electrical insulation of a leaking coil because, unlike water, the oil will not dissolve electrolytes and, therefore, retains a high dielectric strength as it wets the insulation. The oil cooling system is composed of a pump, heat exchanger and piping, and associated control and protective action circuits. Due to higher viscosity and lower heat capacity than water, oil removes 0.89 MW (maximum) at 12 L/s for each coil (1.4 MW at 12 L/s for water). This decreases the repetition rate to one maximum power pulse every 900 s, as opposed to one every 300 s for water. A thermal analysis is presented, and the equipment and controls are described. A flow diagram for the system is provided. >

The PBX-M passive plate stabilization system provides significant stabilization of long-wavelength external kink modes, the slowing of vertical instability growth rates, and the amelioration of disruption characteristics. The passive plate stabilization system has allowed the use of LHCD and IBW to induce current density and pressure profile modifications, and m=1 divertor biasing for modifying edge plasma transport. Improvements in the passive plate system insulators and support structures have provided reliable operation. Impurity influxes with the close-fitting passive plates are low. Solid target boronization is applied routinely to reduce conditioning time and maintain clean conditions.

A diborane gas handling system has been constructed for the purposes of boronizing TFTR. Numerous safety features have been incorporated to mitigate the toxic and inflammable nature of diborane, including many features adopted from the semiconductor industry. The carborane films deposited in TFTR have significantly improved the operational performance of the machine.