Tritium retention and transport is a critical phenom­enon for graphite in fusion systems in general, and it is the subject of a chapter by Causey, San Marchi, and Karnesky in this series. In the previous section of this

Mean ambient thermal conductivity (W m-1 K-1)

chapter, 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 the use of graphite as a PFM.85 Quantification ofthe problem and determination ofpossible mitigat­ing steps is complicated by experimental data, which can vary by orders of magnitude,86-92 as reviewed by Wilson.93

The primary concern over retention of fuel in the PFC is the inventory of hydrogen adsorbed into the graphite and the subsequent release of near-surface hydrogen (due to physical or chemical sputtering, etc.) as plasma discharge begins. The hydrogen sput­tered from the wall oversupplies the plasma edge with fuel, causing instabilities and making plasma control problematic. Tritium inventory concerns are generally safety-related but can have significant economic consequences because of the high cost of tritium. Tritium release to the environment in an accident situation had limited the allowed inventory in TFTR, and was a significant consideration for the sighting of the ITER. It has been estimated84 that as much as ~1.5 kg of tritium would reside in the graph­ite PFM of ITER, corresponding to an additional fuel cost of 1.5-3 million dollars.

A source of trapped hydrogen, not discussed in detail here, which may dominate the tritium inventory in ITER-like machines, is the ‘codeposited layer.’94 This layer is formed by the simultaneous deposition of carbon, which is eroded from the first wall, and hydrogen. Thick layers of carbon redepos­ited to low erosion areas are common, and have been seen in all large tokamaks utilizing graphite PFMs. As this layer grows, the hydrogen contained therein cannot be liberated by surface sputtering and becomes permanently trapped. This problem is unique to graphite and will require continual sur­face conditioning to minimize the total inventory of trapped species. It represents a vast sink for tritium and therefore must be managed in some way. Below is a discussion of the retention of tritium in bulk graphite.

The physical process involved in the retention of hydrogen, as it corresponds to graphite PFMs, is fairly well understood. The energetic hydrogen iso­topes are implanted to depths of less than a micron in the PFM surface. Once implanted, the hydrogen ions are either trapped, reemitted, or diffuse through the bulk. At temperatures less than 100 °C,95,96
the majority of ions are trapped near the end of their range. These trapped ions are not in solution in the graphite, but are held97 in the highly defected struc­ture. The amount of hydrogen isotope that can be accommodated is largely dependent on the implan­tation temperature,96,98 the trap types and densities (defects), and, to a lesser extent, by the implantation depth.9 , One model for bulk hydrogen trapping presented by Atsumi101 is shown in Figure 28. In his work, two distinct traps have been identified. The lower energy trap (2.6 eV) is associated with edge planes of the graphite crystals, with total trapping therefore depending on the effective size and accessi­bility of the crystals to diffusing hydrogen species. The second, higher energy trap (4.4 eV), is associated with dangling bonds, albeit at the edge of an interstitial loop. As the formation of small interstitial loops is one of the primary effects of neutron irradiation, the formation of these deeper traps is directly affected by neutron fluence. The total retained isotopic H can reach as much as 0.4—0.5 H/C in the implanted layer

95,100,102

at room temperature.

As the amount of implanted hydrogen increases toward its saturation value, a larger fraction of ions are released from the graphite surface. At intermedi­ate and high temperatures (>250 °C), diffusion of hydrogen in the graphite lattice occurs. This diffu­sion is most likely along internal surfaces such as micropores and microcracks, while transgranular dif­fusion has been seen above 750 °C.103,104 This bulk diffusion, along with the associated trapping of hydrogen at defect sites, has been studied widely with quite variable results. This variation can be seen in Figure 29, where the temperature depen­dence of the hydrogen diffusion coefficient for sev­eral carbon and graphite materials is shown.

It is expected that the diffusion of hydrogen through graphite would be highly dependent on the graphite microstructure, which may explain the wide range of the data of Figure 29. In any event, the transport of hydrogen through the bulk graphite and associated solubility limits can significantly increase the hydrogen inventory for fusion devices. The effect of the perfection of graphitic structure on the solu­bility of hydrogen is shown by Atsumi’s data105 in Figure 30. The data in Figure 30 indicate that the more defect-free, highly graphitized materials have a lower solubility limit. Further evidence for the role of structural perfection comes from the observation that materials that have been disordered by neutron irradiation have significantly higher solu­bility for hydrogen.105,106

The effect of atomic displacements on the hydro­gen retention of graphite was first shown by Wampler using 6 MeV ion beams.107 Wampler used four types of intermediate and high quality graphites, which were irradiated with a high-energy carbon beam at room temperature, and then exposed to deuterium gas. Wampler’s results indicated that the residual deuterium concentration increased by more than a factor of 30-600 appm for displacement doses appropriate to ITER. However, for reasons that are not entirely clear yet, neutron-irradiated high-quality CFCs retain significantly less tritium than expected

Absorption (Rate-determining step) (>90%)

Desorption О О Detrapping

Figure 28 Schematic of the processing of hydrogen trapping and diffusion in graphite. Reproduced from Atsumi, H. J. Nucl. Mater. 2003, 313-316, 543-547.

from the earlier work. This was reported by Atsumi105 and clearly shown in the work of Causey108 (Figure 31). Causey irradiated high thermal conduc­tivity MFC-1 unidirectional composite and FMI-222

4.0	on n
IG-430U
| |Trap 2 (crystallite edge surface) I I Trap 1 (interstitial cluster loop)
1.9 X 1024 5.4 X 1024 0.23dpa 0.65dpa
Figure 32 Loading of defects in irradiated graphite with hydrogen. Reproduced from Atsumi, H.; Muhaimin, A.; Tanabe, T.; Shikama, T. J. Nucl. Mater. 2009, 386-388, 379-382.

magnitude less than that expected from the earlier work on GraohNOL-N3M.109

In some more recent work by Atsumi,110 different neutron-irradiated graphites were irradiated and the amount of hydrogen that could be entrained in the crystal was measured. Specifically, Atsumi110 irra­diated three isomolded grades of graphite (IG-110, IG-430U, and IG-880U) in the JMTR reactor below 200 °C to various fluences and then baked the sam­ples in an atmosphere of hydrogen. Figure 32 shows the relative abundance of hydrogen that can be loaded into the crystallite edge and interstitial cluster loop — type defects (Type 1 and Type 2 defects of Figure 28) of the IG-430U graphite. Clearly, both defects are produced during irradiation and are accessible to postirradiation loading of hydrogen. In the same work, Atsumi carried out a series of preloading anneal­ing of samples, which suggested that the edge-type defects would be preferentially annihilated. In this context, it is important to note that all work on hydrogen or tritium retention in irradiated graphite has followed the approach of irradiating the material at a relatively low temperature and then loading and unloading and measuring the released hydrogen from the sample at a comparatively higher tem­perature. This may be of significance in that, as inferred in the work of Atsumi and others,105,106,108 the relative crystalline perfection (amount of intrin­sic defects) is strongly related to hydrogen retention. As discussed in Section 4.18.3 and Chapter 4.10, Radiation Effects in Graphite, irradiation at low temperature may result in a significantly different microstructure (an abundance of simple interstitial and vacancy clusters) as compared to the more fusion
relevant irradiation temperatures (formation of more perfect interstitial discs and collapsing of vacancy complexes). Moreover, the postirradiation annealing of a low-temperature-irradiated microstructure will likely not produce representative microstructures of irradiation at more relevant higher temperatures. For this reason, data generated to date should be consid­ered as a guide for the trends likely to occur rather than as quantitative information on the actual tritium retention that will occur in fusion devices. Moreover, they are likely overly conservative.