Recent experiments on the Tokamak Fusion Test Reactor have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition, and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) transiently increased to as high as 75%. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean-up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing ≊8000 Ci and restoring the tritium inventory to a level well below the administrative limit.

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Recent observations have indicated that lithium pellet injection wall conditioning plays an important role in achieving the enhanced supershot regime in TFTR. However, little is understood about the behavior of lithium-coated limiter walls, interacting with edge plasmas. In the final campaign of TFTR, a cylindrical carbon fiber composite probe was inserted into the boundary plasma region and exposed to ohmically-heated deuterium discharges with lithium pellet injection. The ion-drift side probe surface exhibits a sign of codeposition of lithium, carbon, oxygen, and deuterium, whereas the electron side essentially indicates hgh-temperature erosion. It is found that lithium is incorporated in these codeposits in the form of oxide at the concentration of a few percent. In the electron side, lithium has been found to penetrate deeply into the probe material, presumably via rapid diffusion through interplane spaces in the graphite crystalline. Though it is not conclusive, materials mixing in the carbon and lithium system appears to be a key process in successhl lithium wall conditioning.