Tritium profile in plasma-facing components following D–D operation

We have investigated the tritium depth profile near the surface of the limiter/divertor tiles used in the deuterium fueled machines, such as TEXTOR, TFTR and JT-60U by means of the imaging plate technique and a tritium survey monitor. Tritium depth profiles near the surface of the sample tiles were estimated by comparing the experimental results to a calculation using a 3-D Monte-Carlo code. In every sample tile, there was little tritium in the range from the surface to 1 μm depth. In contrast, tritium density tended to increase beyond 1 μm depth. These results indicate that the tritium retained near the surface was easily removed by isotope exchange with a deuterium plasma or various other tritium removal operations. On the other hand, such operations did not remove tritium retained beyond 1 μm depth, and this could be a potential issue in a next D–T machine.