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TFTR performance has surpassed many of the previous tokamak records. This has been made possible by the use of tritium as fuel for DT plasma discharges. Stable operations of tritium systems provide for safe, routine DT operation of TFTR. In the preparation for DT operation, in the commissioning of the tritium systems and in the operation of the Nuclear Facility several key lessons have been learned. They include: the facility must take the lead in interpreting the applicable regulations and orders and then seek regulator approval; the use of ultra high vacuum technology in tritium system design and construction simplifies and enhances operations and maintenance; and central facility control under a single supervisory position is crucial to safely orchestrate operational and maintenance activities.

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980`s The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America`s first operational D-T fusion reactor. This paper will discuss negative pressure employing `elephant trunks` in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and {Delta} pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs.

During 2 yr of experimental operations, the Poloidal Divertor Experiment (PDX) inner wall neutral beam graphite armor provided protection for perpendicular heating injections into normal and disruptive plasmas as well as injection in the absence of plasma for special experiments, calibrations, and tests involving the optimization and development of the PDX neutral beam injection system. About 80 to 100 heating injections occurred per operating day, at a 360-s duty cycle, into plasmas of various densities, and typically approx. =5 to 50% of the injected neutral beam power was transmitted to the armor. More than 10/sup 3/ neutral beam pulses of 100- to 300-ms duration were injected in the absence of plasma at peak power densities of 1.5 to 3 kW/cm/sup 2/, yielding peak surface temperatures of 950 to 1550/sup 0/C. There was no significant impurity production attributable to beam heating of the armor, and no observed beam-induced, macroscopic surface damage. Many of the design constraints and performance issues encountered in this work are relevant to the design of larger fusion devices.