Beryllium is a low-Z material with good thermal char­acteristics, described in Chapter 2.11, Neutron Reflector Materials (Be, Hydrides) and Chapter 4.19, Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices. Additionally, it is a good getter for oxygen impurities in the plasma. The low-Z minimizes the radiation losses from the plasma, and the oxygen removal keeps the plasma clean. For these reasons, beryllium has been used in the JET fusion reactor and will be the first wall material for the International Thermonuclear Experimental Reactor (ITER). Beryllium has interesting hydrogen retention behavior. Beryllium may also be used as a neutron multiplier in the blanket area of future fusion devices to increase the tritium breeding ratio.

Abramov et a/.72 used two grades of beryllium in their permeation-diffusion experiments. These were high-purity (99%) and extra grade (99.8%). Adding to the validity of their experimental result was the fact that the authors used multilayer permeation theory analysis to take permeation through the outer oxide layer into consideration. For a lower
purity material (98%), Tazhibaeva et a/.73 also used the multilayer permeation analysis to determine dif — fusivity. Jones and Gibson74 studied tritium diffusiv — ity and solubility for arc-cast beryllium in the temperature range of 673-1173 K. Beryllium was exposed to tritium gas for various temperatures, durations, and pressures during isothermal anneals. After removing the samples to another experimental system, the samples were heated to various tempera­tures. For the initial heating, the tritium release would rise, but soon fall to zero. Elevating the tem­perature would reestablish the tritium release, but again the release would fall. While this behavior is not typical of diffusion controlled release, the data were analyzed to extract an effective diffusivity. The different reported diffusivities are shown in Figure 8. It can be seen that the diffusivity reported by Abramov et a/.72 is considerably larger than those of Tazhibaeva et a/.73 and Jones and Gibson.74 It is apparent that the purity of the beryllium played a strong role in deter­mining the effective diffusivity. Oxygen, the primary impurity in beryllium provides a strong trap for hydrogen. Thompson and Macaulay-Newcombe75,76 examined the diffusion of deuterium in single-crystal and polycrystalline beryllium. The effective diffusivity in the single-crystal material was lower than that for the polycrystalline material. The polycrystalline results agreed quite well with the results reported by Abramov et a/.72 They suggested that the lower diffusiv — ity seen for the single-crystal samples was the true diffusivity for beryllium, and that the polycrystalline results represented diffusion along the grain boundaries.

Temperature, 1000/ T (K-1)

If hydrogen isotopes migrate along the grain bound­aries, it is logical that the rate of migration would be affected by oxygen segregated to those boundaries.

The very limited results for hydrogen isotope solubility in beryllium are shown in Figure 9. In the earlier described experiments by Jones and Gibson,74 the solubility was seen to be effectively independent of temperature in the temperature range 550-1250 K. For sintered, distilled a-beryllium, Shapovalov and Dukel’skii77 reported similar values of solubility for the temperature range 673-1473 K. In experiments using 98.5% and 99.8% pure beryl­lium samples, Swansiger78 used gaseous uptake of tritium to determine the solubility. The amount of tritium uptake did not increase with increasing sample size. The solubility for the two purity materi­als was also seen to be the same. For temperatures below 650 K, the apparent solubility increased; this strange effect was attributed to trapping. It is interesting to note the fact that the apparent
solubility over temperatures at which the three research groups‘ ’ ’ performed their experiments varied by less than one order of magnitude even though the activation energies varied by 96 kJ mol-1. It should be questioned whether the reported values really represent the solubility of hydrogen isotopes in bulk beryllium.

For all plasma-facing materials, there is concern that implantation of energetic deuterium and tritium could lead to excessive retention and permeation. Implantation of hydrogen isotopes into a material with a low recombination-rate constant can lead to a majority of the hydrogen being pushed into the bulk of the material. In the limiting case of slow recombi­nation, 50% of the hydrogen exits the front face and 50% exits the rear face. Langley79 implanted 25 keV deuterium into 99.1% pure hot isostatic pressed beryllium. The retention was seen to be 100% until the particle fluence reached 2 x 1022 D m — . The retention flattened to a limit of ^2.8 x 1022 D m-2. Wampler80 recorded similar results for his implanta­tion of 0.5 and 1.5 keV deuterium into 99.6% pure beryllium samples. Saturation occurred at 0.31 D/Be in the implant zone. Yoshida et a/.81 used 99% pure beryllium in his implantation experiments with 8 keV deuterons. Transmission electron microscopy revealed bubble formation at all temperatures between room temperature and 873 K. The bubbles were not removed even by annealing at temperatures up to 973 K. Plasma exposure was used by Causey et a/.82 and by Doerner et a/.83 in low-energy, high-fluence deuterium exposures to beryllium. In both studies, the fractional retention was extremely low and decreased with increasing temperature. Open poros­ity in the implant zone was listed as the likely cause of the low retention. Chernikov et a/.84 and Alimov et a/.85 showed bubbles and microchannels to be responsible for the behavior of implanted hydrogen in beryllium. At 300 K, very small bubbles with a high volume density are formed even at low fluences. As the fluence is increased, the bubbles agglomerate into larger bubbles and then form microchannels that eventually intersect with the surface. For irradiation at 500-700 K, small facetted bubbles and large oblate, gas-filled cavities are formed. This microstructure was seen to extend well beyond the implant zone. Alimov et a/.85 postulated that the hydrogen retention in the porous region was due to binding to the beryl­lium oxide that forms on the pore surfaces.

Beryllium is known as a neutron multiplier because of the reaction 9Be + n! 8Be + 2n. Another neutron reaction for beryllium is 9Be + n! 4He + 6He, followed by 6He decaying to 6Li. 6Li has a very large cross-section to absorb a thermal neutron and produce a helium atom and a tritium atom. Baldwin and Billone86 calculated the amount of tritium that could be produced in a large fusion device of the future. In an experiment, they exposed beryllium to a neutron fluence of 5 x 1026 n m-2 with 6% of the neutrons having energy > 1 MeV. The resulting tri­tium level was determined to be 2530 appm. Scaling up to a fusion reactor with 50 Mg of beryllium exposed to 3 MWy m-2 results in the production of

5.5 kg of tritium. This is a sizeable quantity of tritium. The relevant question is whether this tritium would be released during normal operation of the fusion plant. Baldwin and Billone86 examined exactly that question in their experiment. The samples containing the 2530 appm of tritium were heated in stepped anneals to determine the release rate of tritium from beryllium materials with different densities. The annealing began with a very long anneal at 773 K, and the temperature was increased in incre­ments of 100 K. For each temperature, there was a nondiffusional burst of release followed by a rapid decrease in the release rate. The release behavior for the different materials was similar, but the fractional release was greater for the less dense materials. Andreev et a/.87 irradiated hot-pressed beryllium at 373 K. After neutron irradiation, thermal desorption spectroscopy was used with a heating rate of10 K s-1. Release began to occur at -~773 K. The temperature at which maximum release occurred depended on the neutron fluence. The sample irradiated to a fluence of 3 x 1025nm-2 had a peak release at 1080 K, while the sample irradiated to the higher fluence of 1 x 1026nm-2 exhibited a peak release at a lower temperature of 1030 K. The authors examined the microstructure of the samples after the release anneals. If the anneal was stopped at 973 K, pores with a diameter of 2-16 pm were formed. If the anneal was taken to 1373 K, the pore diameters increased to 25-30 pm.

Due to the toxicity of beryllium, there have been relatively small numbers of experiments performed on the behavior of hydrogen isotopes in beryllium. The apparent diffusion coefficient of hydrogen in beryllium is strongly affected by purity levels. The values determined for the solubility of hydrogen in beryllium all fall within one order of magnitude even though the apparent activation energy differs by 96 kJ mol — . Implantation of hydrogen into beryllium results in the formation of bubbles and eventually open channels or porosity. Connection of the porosity to the surface facilitates the release of hydrogen from the beryllium as the particle fluence is increased. The tendency to form bubbles would suggest that the solubility of hydrogen in beryllium is extremely small. It is possible that the values determined for the solubility of hydrogen in beryl­lium actually represent the amount of hydrogen absorbed on the external surface and on the grain boundaries. The measured diffusivity may represent migration along the grain boundaries. More experi­ments, and experiments with single crystals, are needed to answer these questions. For beryllium used for long times in future fusion devices, tritium produced by neutron reactions on the beryllium is likely to dominate tritium retention in beryllium. Tritium inventory from eroded beryllium codepos­ited with tritium may play a strong role in tritium inventory, but that effect is not covered in this review.