Tritium retention in graphite

In the previous section the interaction of the plasma particle flux with the surface of graphite was discussed. However, the fate of the implanted particles (most importantly deuterium and tritium) following their impact with the graphite surface is also an important issue, and is seen by some as the major impediment to graphite s use as a PFM [58], Quantification of the problem, and determination of possible mitigating steps, is complicated by experimental data which can vary by orders of magnitude [59-66] as reviewed by Wilson [67]. 

Next generation machines will impose increasingly greater thermal loads on their PFCs. High thermal conductivity CFC materials may offer a solution to the high-heat loads, but further research is needed to overcome the problems noted above and to assure the place of carbon materials in future fusion power reactors. 

Research sponsored by the U.S. Department of Energy under contract DE-AC05-960R22464 with Lockheed Martin Energy Research Corporation at Oak Ridge National Laboratory. 

Watson, et al., High Heat Flux Testing of CITFirst Wall Tiles, 1988. 

Simmons, Radiation Damage in Graphite, Pergamon Press, (1965 ). 

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Fig. 18. Tritium retention as a function of neutron damage in graphite and graphite composite.

There are large uncertainties about the tritium retention problem. The most pessimistic extrapolations based on data from co-deposition of tritium with carbon [8] conclude that the allowed maximum of 350 g of T retained in the vessel will be reached after only a few ITER discharges. This is seen as a genuine problem with graphite, whereas other candidate wall materials like W do not show such a strong effect of tritium retention. Therefore, already for ITER it is of paramount importance to clarify this problem and to verify that carbon as a wall material is acceptable at all for tritium operation. 

Fig. 10.4. Deuterium retention in POCO AXF-5Q graphite as measured by tritium...

Plasma conditions and wall materials must also enable a sufficient lifetime of the first wall components for economic reasons. Chemical erosion of graphite leads to significant erosion yields even under low-temperature, cold plasma conditions and can seriously limit the lifetime. Since the tokamak is a fairly closed system, most of the eroded material will be re-deposited somewhere inside the machine. The question of tritium retention and overall inventory in the device is closely connected to the chemical erosion and to possible co-deposition as well [6,7]. In order to minimize the net-erosion and optimize the lifetime of wall components, the re-deposition should be concentrated in areas of major erosion. Another way to minimize chemical erosion is the use of mixed materials, which - in laboratory experiments - display a reduced erosion yield in comparison to pure graphite. 

Graphite itself also has the ability to bind and contain tritium, and some modern efforts seek to employ that behavior through increased surface area such as the case with pebble bed, molten salt-cooled designs (Peterson et al., 2008). Graphite also has been seen to increase its tritium retention with neutron irradiation, which aids in this area (Causey, 1989). 

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