4.19.3.2.1 Implantation

The use of beryllium as a plasma-facing material in tokamaks has prompted many experimental studies of retention and emission of hydrogen implanted into beryllium-like metals from ion-beams or plasmas. References and discussions of these studies can be found in reviews.82- 5 Here, we review those studies which are relevant to H retention in Be in a fusion plasma environment. This section is mainly excerpted from Federici et all Two basic parameters for under­standing H retention are the hydrogen diffusivity and
solubility. Studies of solubility and diffusivity are reviewed in Causey and Venhaus85 and Serra et al.86 Figures 487-90 and 587,91,92 show experimental values for hydrogen solubility and diffusivity in W and Be. For Be there are significant differences between results from various studies. These differences may be due to effects of traps and surface oxide layers. The presence of bulk traps tends to increase the measured values of solubility and to decrease the mea­sured values of diffusivity (see Federici et al.7), especially under conditions where the concentration

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of hydrogen in solution is smaller than the concen­tration of traps. For this reason, studies done on materials of higher purity and crystalline perfection, and at higher temperatures and with higher concen­trations of hydrogen in solution, tend to give more reliable results. The porosity and oxide inclusions present in beryllium produced by powder metallurgy are also likely to lead to inconsistent results in mea­surements of hydrogen solubility and diffusivity. In the Be experiments, the effects of traps were not characterized and may be dominant. One firm con­clusion is that the solubility of hydrogen is very low in both Be and W.

Many studies have been done on the retention and emission of H implanted into materials to provide data needed to predict H retention in fusion reactor environments. Figure 6 shows the retention of 1 keV deuterium implanted into Be at 300 K versus inci­dent fluence, measured by thermal desorption.9 D retention in Be was close to 100% at low fluences but saturated at high fluences. Earlier nuclear reac­tion analysis (NRA) measurements of D retained in Be within ~1 pm of the surface gave very similar results.94 This saturation behavior indicates that D implanted into Be at 300 K does not diffuse, but accumulates until it reaches a limiting concentration of ^0.3-0.4 D/Be within the implantation zone. At high fluences, the implanted zone becomes porous allowing additional implanted D to escape. This
saturation mechanism is confirmed by electron micros­copy, which shows bubbles and porosity in the implan­tation zone after high fluence H implantation.95 Saturation of retention by the same mechanism is observed for D implanted into stainless steel at 150 K where the D is not mobile.96 H retention in Be increases with increasing ion energy and decreases with increasing sample temperature.84,97 The retention of 1 keV deuterium implanted into W and Mo at 300 K98 is also shown for comparison in Figure 6.

Figure 784 shows retention of deuterium and tri­tium as a function of incident particle fluence from a set of high fluence experiments in which Be specimens were exposed to laboratory ion-beams (Idaho National Engineering and Environmental Laboratory (INEEL), University of Toronto Institute for Aerospace Studies (UTIAS)), linear plasma devices (Sandia National Laboratory (SNL)/Los Alamos National Laboratory — Tritium plasma experiment (LANL-TPE), University of California, San Diego-Plasma Interaction with Surface Components Experimental Station B (UCSD — PISCES-B)), a tokamak divertor plasma (DIII — D-DIMES), and a neutral beam (NB-JET). In some of these studies carbon deposition or formation of carbide or oxide surface layers occurred, which is likely to affect D retention. The figure shows the D retention in Be observed under a wide range of exposure conditions. The high fluence saturated con­centration tends to be lower at higher temperatures.

It must be noted that this phenomenon is very important because it implies that tritium inventories and permeation due to implantation in beryllium for ITER PFC applications should be significantly lower than was previously estimated using classical recombination-limited release at the plasma surface. A first attempt to model this saturation by allowing the recombination coefficient to become exponen­tially large as the mobile atom concentration near the plasma-facing surface approaches a critical value was made by Longhurst et al.99 For Be, calcula­tions suggest that the critical concentration is related to the yield strength using Sieverts’ law of solubility. On the basis of the results of these calculations, it can be concluded that the inventory of tritium in the beryllium first wall of a device such as ITER, because of implantation, diffusion, trapping, and neutron — induced transmutation, will be of the order of 100 g rather than the kilogram quantities estimated previ­ously,100,101 and most of that will result from neutron — induced transmutations in the Be itself and from trapping in neutron-induced traps. Current predic­tions of tritium inventory in ITER are briefly dis­cussed in Section 4.19.6.2.1.

Fusion neutrons will create vacancies and intersti­tials in plasma-facing materials. For metals at reactor wall temperatures, these defects will be mobile and will annihilate at sinks (e. g., surfaces or grain bound­aries), recombine, or agglomerate into defect clusters. Vacancy agglomeration may also lead to the formation
of voids. In beryllium, neutron-induced nuclear reac­tions produce helium and tritium, which may be trapped at defects or precipitate as gas bubbles. These defects, resulting from neutron irradiation, will increase the retention of hydrogen, by increasing the concentration of sites where diffusing hydrogen can precipitate as gas or become trapped as atoms. The effect of neutron irradiation on hydrogen reten­tion in metals is complex, but, in principle, this can be modeled, provided the material parameters are known, such as hydrogen diffusivity, solubility, trap binding energy, and defect microstructure produced by the neutron irradiation. For many metals, most of these parameters are known well enough to attempt such modeling. For beryllium, however, uncertainties in solubility, diffusivity, and trapping of hydrogen make such modeling of hydrogen retention difficult.

The problem of production of helium and tritium by nuclear transmutation in beryllium itself is dis­cussed in Section 4.19.4.4.5.