4.16.3.1 Plasma-Facing Materials

Tritium generated in the fusion-reactor blanket will be fed directly into the plasma in the main vacuum cham­ber. There, the tritium will be partially consumed, but it will also interact with the materials composing the first wall. Materials used to line the first wall will be exposed to energetic tritium and deuterium escaping from the plasma. Particle fluxes in the range of 1021 (D + T) m-2 s-1 will continuously bombard the plasma-facing materials. Materials used for the diver­tor at the top and/or bottom of the torus will be exposed to lower energy particles with a flux of 1023 (D + T) m-2 s-1 or higher. While the neutral gas pressure of tritium will be relatively low at the outer vacuum wall boundary, some minor permeation losses will occur. In reality, the primary concerns in the plasma-facing areas are tritium inventory and perme­ation into the coolant through coolant tubes. While future power reactors are likely to have primarily refractory metals such as tungsten, present-day devices are still using carbon and beryllium. In this section on plasma-facing materials, we examine the interaction of tritium with carbon, tungsten, and beryllium.

4.16.3.1.1 Carbon

In many ways, carbon is ideal for fusion applications. It is a low-Z material with a low vapor pressure and excellent thermal properties. The carbon used in fusion applications comes in two forms, graphite and carbon
composites. Graphite is described in Chapter 2.10, Graphite: Properties and Characteristics; Chapter 4.10, Radiation Effects in Graphite, and Chapter 4.18, Carbon as a Fusion Plasma-Facing Material

and is typically made using the Acheson process.33 Calcined coke is crushed, milled, and then sized. The properties of the graphite are determined by the size and shape of these particles. Coal tar is added to the particles and the batch is heated to 1200 K. This process is repeated several times to increase the density of the compact. The final bake is at temperatures between 2900 and 3300 K and takes ~15 days. The final product is quite porous, with a density of around 1.8—1.9 g cm-3 (compared to a theoretical density of 2.3 g cm-3). Graphite is composed of grains (from the original coke particles) with a size of 5-50 pm, which are in turn composed of graphite subgrains with a typical size of 5 nm. Carbon composites are made by pyrolyzing a composite of carbon fibers in an organic matrix. These fibers have a high strength-to-weight ratio and are composed of almost pure carbon. As with graphite, carbon composites are quite porous with a density of <2 g cm — . Carbon (as either graphite or carbon composite) will not serve as a vacuum or coolant boundary. Therefore, permeation through the carbon will not directly affect tritium release into the environment. Coolant tubes inside the graphite will control the tritium release to the coolant system. The ways in which hydrogen isotopes can interact with carbon are explored in the following sections.

It is difficult to consider hydrogen isotope perme­ation in graphite in the same manner as one would for the metals. As was shown by Kiyoshi eta/.,34 hydrogen passes rather readily through graphite in the molec­ular form. It is when one considers the uptake or retention of atomic hydrogen in graphite or a carbon composite that it becomes more interesting. Most graphites have a Brunauer-Emmett-Teller (BET) or specific surface area of 0.25-1.0 m2 g-1.35 This presents a lot of surface area for the absorption of hydrogen isotopes. Barrer,36 Thomas,37 and Bansal38 are only a few of the many who have reported hydro­gen retention on carbon surfaces. Activation energies varying from 24 to 210 kJ mol-1 have been reported. Once absorbed on a carbon surface, the hydrogen can migrate by jumping from one chemically active site to another. In a fusion reactor, this process occurs in carbon at lower temperatures at which the plasma provides ionic and atomic hydrogen for direct absorp­tion. Molecular hydrogen can dissociate on carbon surfaces, but only at elevated temperatures (>1000 K). As a plasma-facing material, graphite will be exposed to atomic tritium and deuterium, and these hydrogen isotopes will migrate inward along the open porosity. Several research groups have measured the diffusion coefficient for hydrogen on carbon surfaces. Robell eta/.39 inferred an activation energy of 164kJ mol-1 for the diffusion during measurements on the uptake of hydrogen on platinized carbon between 573 and 665 K. Olander and Balooch40 used similar experi­ments to determine the diffusion coefficient for hydro­gen on both the basal and prism plane: 6 X 10-9 exp (-7790/7) m2 s-1 for the basal plane and 6 X 10-11 exp(-4420/7) m2s-1 for the prism plane. Causey eta/.41 used tritium profiles in POCO AXF-5Qgraphite exposed to a tritium plasma to extract a diffusivity for tritium on carbon pores of 1.2 X 10-4 exp(-11 670/7) m2 s-1. An example of the deep penetration of hydro­gen isotopes into the porosity of graphite was reported by Penzhorn et a/.42 Graphite and carbon composite tiles removed from the Joint European Torus (JET) fusion reactor were mechanically sectioned. The sec­tions were then oxidized, and tritiated water was collected for liquid scintillation counting. Relatively high concentrations of tritium were detected tens of millimeters deep into the tiles.

The diffusion or migration of hydrogen isotopes on carbon surfaces occurs at lower temperatures. The solubility, diffusion, and trapping of hydrogen in the carbon grains are higher temperature processes than adsorption and surface diffusion. At higher tempera­tures (>1000 K), hydrogen molecules can dissociate and be absorbed at chemically active sites on carbon surfaces. Some of these sites are located on the out­side of the grains, but many exist along the edges of the subgrains that make up the larger grains. Hydrogen isotopes dissociating on the outer grain boundary are able to migrate along the subgrain boundaries, entering into the interior of the grain. It is the jumping from one moderate energy site (^240 kJ mol-1) to another that determines the effec­tive diffusion coefficient. Traps on the grain bound­aries pose a binding energy barrier (~175 kJ mol-1) that must be overcome in addition to this normal lattice activation. Atsumi et a/.43 used the pressure change in a constant volume to determine the solu­bility of deuterium in ISO 88 graphite. They deter­mined the solubility to be given by K = 18.9 exp (+2320/7) mol H2 m-3 MPa-1/2 over the tempera­ture range of 1123-1323 K. This solubility is shown in Figure 3 along with two data points by Causey44 at 1273 and 1473 K. A negative heat of solution is seen in both sets of data, suggesting the formation of a bond between hydrogen and carbon.

10

0.65 0.7 0.75 0.8 0.85 0.9

Temperature, 1000/T (K-1)

10-11

10-13

‘co

10-15

>>

10-17 h

10-19 10-21

0.4 0.6 0.8 1 1.2 1.4

Temperature, 1000/T (K-1)

Figure 4 Diffusivity of hydrogen in graphite. The bold line is the best estimate given in Causey44 based on uptake data for tritium in POCO AFX-5Q graphite. Adapted from Atsumi, H.; Tokura, S.; Miyake, M. J. Nucl. Mater. 1988, 155, 241-245; Causey, R. A. J. Nucl. Mater. 1989, 162, 151-161; Rohrig, H. D.; Fischer, P. G.; Hecker, R. J. Am. Ceram. Soc. 1976, 59, 316-320; Causey, R. A.; Elleman, T. S.; Verghese, K. Carbon 1979, 17, 323-328; Malka, V.; Rohrig, H. D.; Hecker, R.

Int. J. Appl. Radiat. Isot. 1980, 31, 469.

The variation in the diffusivities of hydrogen in graphite determined by various researchers is extreme. This variation results primarily from differ­ences in interpretation of the mechanism of diffusion (e. g., bulk diffusion or grain boundary diffusion). Representative values for the diffusion is shown in Figure 4. Rohrig et a/.45 determined their diffusion coefficient using the release rate of tritium from nuclear grade graphite during isothermal anneals. They correctly used the grain size as the real diffusion
distance. Causey et a/46 measured the release rate of tritium recoil injected into pyrolytic carbon to deter­mine the diffusivity. Malka eta/47 used the release rate of lithium-bred tritium in a nuclear graphite to deter­mine the diffusion coefficient. Atsumi et a/43 used the desorption rate of deuterium gas from graphite samples that had been exposed to gas at elevated tem­peratures to determine a diffusivity. Building on the work of others, Causey44 proposed an alternative expres­sion for the diffusivity that he labeled ‘best estimate.’

The result was based on uptake experiments for tri­tium into POCO AFX-5Qgraphite, and the assump­tion that the total uptake is determined by the product of the diffusivity and the solubility. The uptake data were analyzed assuming that the expression given by Atsumi et a/.43 for the solubility was correct. The expression ‘best estimate’ was used because it prop­erly took into consideration that the grain size was the effective diffusion distance, and that diffusivity was more properly determined by uptake than release. Release rates are strongly affected by trapping.

The trapping of hydrogen isotopes at natural and radiation-induced traps has been examined by sev­eral research groups.41,48-50 Causey et a/.41 exposed POCO-AFX-5Q graphite to a deuterium/tritium mixture at elevated temperatures in an examination of the kinetics of hydrogen uptake. For temperatures above 1500 K, it was discovered that increasing the pressure did not increase the retention. It appeared that the solubility hit an upper limit at 17 appm. Anal­ysis of the data revealed that the 17 appm did not represent solubility, but a trap density. The trap was determined to have a binding energy of 175 kJ mol — . Atsumi et a/.48 found that radiation damage increased the apparent solubility of deuterium in graphite by a factor of 20-50 with saturation at a radiation damage level of 0.3 dpa. A significant decrease in the apparent diffusivity was also noted. In a later study, Atsumi et a/.49 reported graphites and composites to vary sig­nificantly in their natural retention values. They saw that the saturation retention was inversely proportional to the lattice constant (which relates to the degree of graphitization, and so grain size). Radiation damage was seen to decrease the apparent lattice constant and increase the saturation retention. 6 MeV C+ ions were used by Wampler eta/.50 to simulate neutron damage to different graphites. The trap density increased with damage levels up to 0.04 dpa at which the saturation retention was 650 appm. In a defining set of experi­ments, Causey eta/.28 examined tritium uptake in unir­radiated and radiated pitch-based carbon composites. Pitch-based carbon composites have a large lattice parameter due to the sheet-like configuration of the fibers. These composites retained significantly less tri­tium before and after irradiation than other carbon materials. From these results, it is apparent that high — energy trapping occurs at the edges of the hexagonal crystals on the prism plane.

Hydrogen isotope permeation in the normal sense does not apply to graphite and carbon composites. There is inward migration of atomic hydrogen isotopes along porosity at lower temperatures. To calculate the potential tritium inventory for this process, one can obtain an upper bound by assuming monolayer coverage of the pore surfaces. Typical nuclear grade graphite has a specific surface area of 1 m2 g-1. Complete loading of that amount of surface area yields 2 x 1025 Tm-3, or about 2 g for a 20 m2 carbon wall that is 10-mm thick. At higher temperatures, molecular hydrogen isotopes, which are moving through the graphite pore system, are able to disso­ciate and enter the multimicron-sized graphite grains. This migration into the grains occurs along the edges of the nanometer scale subgrains. As the hydrogen migrates inward, it decorates high-energy trap sites. The density of these trap sites is higher than the effective solubility derived from the migra­tion rate. Permeation into graphite will not lead to tritium release from a fusion device, but will affect tritium inventory. If one assumes that radiation dam­age from neutrons has increased the concentration of traps with a binding energy of 175 kJ mol-1 up to 1000 appm, and that tritium is occupying 100% of those traps, the same 20 m2 wall 10-mm thick listed above would now contain an additional 100 g of tritium. Occupation of all of the traps is difficult to achieve: at low temperatures, kinetics makes it impos­sible to achieve saturation, while at substantially higher temperatures, the traps do not remain filled.

There is another process called carbon codeposi­tion that can strongly affect tritium inventory in a fusion device. In the codeposition process, carbon eroded from the walls of the tokamak is redeposited in cooler areas along with deuterium and tritium from the plasma. Because carbon codeposition is not a diffusion or permeation process, it will not be covered in this review. The interested reader is referred to a review of this process by Jacob.51