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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.

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 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) operated for 15 years, from 1982 to 1997. From November 1993 to April 1997 a mixture of deuterium-tritium (D-T) was used to fuel experiments, leaving the reactor system in an activated and contaminated state. In this condition, the removal of the TFTR and its subsystems is greatly complicated. In 1999, PPPL commenced the TFTR Decontamination and Decommissioning (D&D) Project. The objectives of the D&D projects are to completely remove the TFTR in a safe, efficient and cost effective manner. To date all diagnostics, structural components and ancillary systems have been removed, and the tokamak is in the process of being sectioned into smaller pieces. This paper will describe the criteria, the methods employed and the successes of the D&D Project.

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. >

Recent experimental results in the fusion community have pointed to possible directions for improvement of the tokamak concept. A critical issue for many advanced tokamak scenarios is the optimization of the spatial distribution of plasma current density. Recent experiments on the Tokamak Fusion Test Reactor and Doublet III-D have demonstrated enhanced confinement and improved stability using transient techniques for the modification of the plasma current profile. The Princeton Beta Experiment-Modification (PBX-M) has begun experiments using profile modification techniques that are extrapolable to long-pulse or steady-state operation. PBX-M uses lower hybrid current drive for current profile modification, ion Bernstein wave heating for pressure profile variation, and plasma shaping to alter the magnetic geometry. This paper is an overview of the PBX-M program, its engineering facilities, and discussion of how this program is relevant to future advanced tokamak programs such as the Tokamak Physics Experiment.

PBX-M upgrades have been in progress for enhanced capability and reliability. These include upgrades to the vessel O-ring system for impurity control, the Ion Bernstein Wave and Lower Hybrid Current Drive Systems for plasma profile control, a double viewing upper and lower CHERS based Poloidal Rotation Diagnostic, and a new digital plasma control system.

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.