The mechanical integrity of nonhydriding metals in the presence of tritium is excellent because the electron bands carry away the energy of colliding beta particles without disrupting the metal structure or bonding. These metals form the most common class of tritium containment structural materials. They generally include 304L, 316L, 321, 21-6-9, and Nitronic stainless steels, as well as copper and aluminum. Inconel, Ni-Cr alloys, and 400-series stainless steels are generally not chosen because of corrosion or hydrogen embrittlement sensitivity. At high pressures of tritium gas, however, classical hydrogen embrittlement, as well as helium-3 embrittlement, can occur in accepted materials. For example, for 304L stainless steel samples exposed to 9 kpsi of tritium at 423 K for 6 months and then aged 1.5 years, fracture toughness decreased by a factor of 6. Of this, a factor of two could be attributed to helium-3 alone.

Substantially different fracture modes are observed between aged tritium-loaded and unloaded steel specimens. Helium-3 is vastly less soluble in metals than is hydrogen (tritium); helium pockets (bubbles) form with high internal pressures. Hydrogen embrittlement also contributes to this effect.

Permeative escape rates of tritium through nonhydriding metals are generally acceptable at temperatures below 100C to 300C and for thicknesses of 0.1 cm or more. For 304 stainless steel 0.3 cm thick with a 1000-CM² surface area exposed on one side to tritium gas of 1 atm pressure at 300 K, the permeability is 1.6 x 10^-4 Ci/day (t_0.9 = 7 hours). The temperature dependence of permeation is often astounding.

Cross-contamination between nonhydriding metals and tritium does occur often enough to be troublesome. Oxide layers on metals often contain hydrogen and are further covered with a thin adsorbed carbonaceous film when originally grown in room air. Upon exposure to such a surface, tritium gas may become contaminated over hours or days with hundreds to thousands of parts per million of protium (as HT) and methane (as CT₄) as the surface layers are radiolyzed, exchanged, and contaminated by the material. Because diffusion of tritium in the bulk material is usually slow at room temperature, the extent of surface oxide contamination may greatly surpass the bulk contamination of a component. Cross-contamination can be minimized by minimizing material surface areas, choosing an impermeable material with a thin or nonexistent oxide layer, and maintaining cleanliness.

Tritium present in an oxide layer can be removed by acid dissolution of the oxide or more gently by isotopic exchange with normal water or activated hydrogen gas (plasma). Because diffusion of oxide- or bulk-dissolved tritium back to the surface of a material undergoing decontamination is often slow, exchange at an elevated temperature may be advantageous.

Hydridin~ Metals

When exposed to tritium gas, hydriding metals absorb large volumes of tritium to form tritide phases, which are new chemical compounds, such as UT₃. The mechanical integrity of the original metallic mass is often severely degraded as the inclusions of a brittle, salt-like hydride form within the mass. Because of this property and their large permeability to hydrogen, hydriding metals are not to be used for constructing pipelines and vessels of containment for tritium gas. They have great utility, however, in the controlled solidification and storage of tritium gas, as well as in its pumping, transfer, and compression.

Uranium, palladium, and alloys of zirconium, lanthanum, vanadium, and titanium are presently used or are proposed for pumping and controlled delivery of tritium gas. Several of these alloys are in use in the commercial sector for hydrogen pumping, storage, and release applications. Gaseous overpressure above a hydride (tritide) phase varies markedly with temperature; control of temperature is thus the only requirement for swings between pumping and compressing the gas.

In practice, pumping speeds or gaseous delivery rates (the kinetic approach to equilibrium) are functions of temperature (diffusion within the material), hydride particle size, and surface areas and conditions. Poisoning of a uranium or zirconium surface occurs when oxygen or nitrogen is admitted and chemically combines to form surface barriers to hydrogen permeation. In practice, these impurities may be diffused into the metal bulk at elevated temperature, thereby reopening active sites and recovering much of the lost kinetics. Other metals and alloys (for example, LaNi₃) are less subject to poisoning, although alloy decomposition can occur.

Helium-3, generated as microscopic bubbles within the lattice of tritides, is not released except by fracture and deformation of metal grains. This release usually occurs at high temperature or after long periods of time. When a tritide is heated to release tritium, helium-3 is also released to some extent. The cooled metal, however, does not resorb the helium-3. The practice of regenerating a tritide storage bed to remove helium-3 immediately prior to use for pure tritium delivery is therefore common.

If helium-3 (or another inert impurity) accompanies tritium gas that is absorbed onto a tritide former, helium blanketing may occur. The absorption rate slows as the concentration of helium in the metal crevices leading toward active sites becomes high. Normal gaseous diffusion is often not sufficient to overcome this effect. Forced diffusion by recirculating the gas supply can be used to overcome blanketing.