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 mum depth. In contrast, tritium density tended to increase beyond 1 mum 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 mum depth, and this could be a potential issue in a next D-T machine.

Subject headings

NATURVETENSKAP  -- Fysik (hsv//swe)

NATURAL SCIENCES  -- Physical Sciences (hsv//eng)

Keyword

imaging plate technique

co-deposited layers

w-shaped divertor

jt-60u

jet

tftr

deuterium

retention

hydrogen

surface

Publication and Content Type

ref (subject category)

art (subject category)