Publications

The experimental data on sorption and solubility of hydrogen isotopes in graphite in a wide ranges of temperature and pressure are reviewed. The Langmuir type adsorption is proposed for the hydrogen -- graphites interaction with taking into account dangling sp{sup 2}{minus}bonds relaxation. Three kinds of traps are proposed: Carbon interstitial loops with the adsorption enthalpy of {minus}4.4 eV/H{sub 2} (Traps l); carbon network edge atoms with the adsorption enthalpy of {minus}2.3 eV/H{sub 2} (Traps 2): Basal planes adsorption sites with enthalpy of +2.43 eV/H{sub 2} (Traps 3). The sorption capacity of every kind of graphite could be described with its own unique set of traps. The number of potential sites for the ``true solubility`` (Traps 3) we assume as 1E+6 appm, or HC=l, but endothermic character of this solubility leads to negligible amount of inventory in comparison with Traps 1 and Traps 2. The irradiation with neutrons or carbon atoms increases the number of Traps 1 and Traps 2. At damage level of {approximately}1 dpa under room temperature irradiation the number of these traps was increased up to 1500 and 5000 appm respectively. Traps 1 and Traps 2 are stable under high temperature annealing.

Three potential methods for measuring the surface tritium content of the TFTR vacuum vessel are described, each based on a different technique for measuring the in situ beta emission from tritium. These methods should be able to provide both a local and a global assessment of the tritium content within the top [approx] 1[mu]m of the inner wall surface.

Experiments were done to determine effect of lattice damage on solubility and transport of deuterium (D) in silicon carbide. Beta SiC samples were irradiated with energetic ions to produce lattice damage, and were then soaked in D{sub 2} gas at 1000C. Concentration of D versus depth was then measured by nuclear reaction analysis. Very near the surface (<0.5{mu}m), concentration of D was larger in irradiated than in unirradiated, but beyond 1 {mu}m the D concentrations were similar ({approximately}20{plus_minus}10 atomic ppM), even though the damage extended to 2.2 {mu}m in most of the samples. Results from this study of ion-irradiated SiC together with our previous study of tritium migration in undamaged SiC point to the conclusion that uptake of D from gas into SiC occurs by transport along grain boundaries, whereas uptake of D into lattice damage produced by ion irradiation, and release of energetically implanted D both require permeation of D within grains which is much slower.

This paper presents the results of an experimental study of the thermal stability in air and vacuum of the codeposited carbon/hydrogen layer formed in a carbon-lined fusion reactor. Results are also presented for the stability of the saturated layer formed by the implantation of energetic hydrogen ions into a graphite surface. For both films, the hydrogen isotope release occurs at a much lower temperature in air than it does in vacuum. At temperatures above 600 K, the hydrogen isotope release in air is very rapid and is emitted in a condensible form. It is speculated that isotopic exchange with water present in the air is responsible for this release. In vacuum, temperatures in excess of 1000 K are required to produce a rapid release from either film. The implications of these results to the safety of tritium in carbon-lined fusion reactors are discussed. © 1990.