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The Advanced Toroidal Facility (ATF) vacuum vessel and its internal components have been conditioned for plasma operation by baking, discharge cleaning with hydrogen and helium, and gettering with chromium and titanium. The plasma‐facing surface of ATF consists mainly of stainless steel with some graphite; the outgassing area is dominated by the graphite because of its open porosity. Since this situation is somewhat different from that in other fusion plasma experiments, in which a single material dominates both the outgassing area and the plasma‐facing area, different cleaning and conditioning techniques are required. The situation was aggravated by air leaks in the vacuum vessel, presumably resulting from baking and from vibration during plasma operation. The results of the various cleaning and conditioning techniques used are presented and compared on the basis of residual gas analysis and plasma performance. A technique for detecting leaks from the inside of the vacuum vessel is described. This technique was developed because access to the outside of the vessel is severely restricted by external components.

Techniques for cleaning and conditioning the vacuum vessel of the Advanced Toroidal Facility (ATF) and its internal components are described. The vacuum vessel cleaning technique combines baking to 150 °C and glow discharges with hydrogen gas. Chromium gettering is used to further condition the system. The major internal components are the anodized aluminum baffles in the Thomson scattering system, a graphite‐shielded ICRF antenna, two graphite limiters, and a diagnostic graphite plate. Three independent heating systems are used to bake some of the major components of the system. The major characteristics used for assessing cleanliness and conditioning progress are the maximum pressure attained during bakeout, the results of gas analysis, and revelant plasma parameters (e.g., time to radiative decay). Details of the various cleaning and conditioning procedures and results are presented.

An experiment has been performed on the Impurity Study Experiment (ISX‐B) tokamak to test beryllium as a limiter material. Beryllium is an attractive candidate for a limiter and has been proposed for use in the Joint European Torus (JET) experiment. A temperature‐controlled, segmented, beryllium top‐rail limiter was located inside the plasma radius described by the existing titanium limiters. An extended set of diagnostics was added for measurement of scrapeoff and limiter parameters. These included visible and infrared monitoring systems, probes, and surface analysis experiments. Tokamak experiments included parameter surveys of both Ohmically heated and neutral‐beam‐heated plasmas and an extended fluence test of the limiter. The most significant effect of operation with a beryllium limiter was the reduction in low‐Z impurities caused by gettering action of beryllium deposited on the liner walls. The experiment required the design and implementation of contamination control apparatus and work procedures ...