R.-D. Penzhorn

Distribution and Mobility of Tritium in Type 316 Stainless Steel

Abstract Exposure of Type 316 stainless steel to tritium-containing hydrogen at an elevated temperature causes diffusion of the majority into the bulk and trapping of a small fraction in a thin oxide layer on the surface at concentrations far exceeding those in the bulk. The uptake by the bulk and surface layer is temperature and pressure dependent. After chemical erosion of the tritium-rich layer, the concentration of tritium on the topmost surface is slowly and asymptotically restored even at 298 K. Isothermal heating at 373 or 473 K until substantial release of the bulk tritium is associated with a comparatively much smaller liberation from the surface layer suggesting different retention and liberation mechanisms. The tritium inventory and profile evolution of homogeneously loaded Type 316 stainless steel caused by chronic release at the ambient temperature and radioactive decay were followed experimentally over several years and modeled successfully by a diffusion mechanism. The model has been adapted to specimens nonhomogeneously loaded with tritium only up to the subsurface. It simulates profile and inventory changes well even after prolonged aging. Chronic release constitutes an aging loss of tritium comparable to that of radioactive decay that should be taken into account for the assessment of tritium-contaminated stainless steel waste.

Tritium Interaction with Surface Layer and Bulk of Type 316 Stainless Steel and Consequences of Aging

Abstract Stainless steel exposed to gaseous tritium characteristically shows a firmly trapped fraction of tritium in the surface layer, which is not fully removable by water at ambient temperature. Prolonged thermal treatment of tritium-loaded specimens at <443 K causes substantial depletion of the bulk but almost no depletion of the surface layer. For complete removal of hydrogen isotopes from the bulk and the surface, temperatures exceeding 573 K are necessary. Upon chemical etching virtually all tritium trapped in the surface layer appears in the etching solution as tritiated water. Following removal of the layer by chemical etching, the tritium-rich layer reappears after months of aging at ambient temperature with nearly the original tritium activity. Comparison of chronic tritium release rates into liquid water before and after etching reveals that the surface layer only marginally influences the rate. X-ray photoelectron spectroscopy provides evidence that during prolonged aging the surface layer continues to grow while at the same time trapping a fraction of bulk tritium released at ambient temperature. Experimental results suggest different mechanisms of hydrogen uptake and release by the bulk and surface layers. Inference of tritium activity in the bulk of aged or heat-exposed stainless steel material from surface activity measurements may depart significantly from reality.

On the Fate of Tritium in Nickel

Abstract Uptake, distribution, and release behavior of tritium in Ni was investigated by chemical etching and thermal release rate measurements. Liberated tritium was found to consist almost exclusively of tritiated water. The chronic release rate of tritium from Ni was significantly larger than that from type 316 stainless steel. Depth profiles in specimens that partially lost tritium due to its chronic release into vacuum, air or a stream of argon could be reproduced by a one-dimensional diffusion model using best fit diffusion coefficients. Values of the best-fit diffusion coefficients at 298 K were found to be independent from the ambient into which tritium was released. The average diffusion coefficient from all measurements at 298 K, i.e. (2.7 ± 1.3) × 10-10 [cm2/s] was in line with diffusion coefficients calculated from literature data at the same temperature. Hence, the diffusion model constitutes a useful tool for the prediction of tritium bulk depth profiles in Ni during chronic release (CR).

Uptake and Distribution of Tritium in Copper

Abstract The uptake of tritium on the surface and in the bulk of copper upon exposure to a 50 % T/H mixture at 300 or 473 K was investigated using a chemical etching technique. Concentrations of tritium approaching saturation are achieved fairly rapidly in Cu even at low temperatures because of comparatively high diffusivity and low solubility of hydrogen in this material. The results were interpreted by a diffusion model. Most notorious are the very high concentrations of tritium on the topmost surface and subsurface. They were quantified by etching and confirmed by BIXS. In addition, there is evidence for tritium trapping in the subsurface region. Tritium-loaded copper specimens release tritium chronically at ambient temperature. The egress of tritium manifests in the gas phase almost exclusively as tritiated water.

Tritium release from SS316 under vacuum condition

Abstract Type 316 stainless steel specimens loaded with tritium either by exposure to 1.2 kPa HT at 573 K or submersion into liquid HTO at 298 K showed characteristic thin surface layers trapping tritium in concentrations far higher than those determined in the bulk. The evolution of the tritium depth profile in the bulk during heating under vacuum was non-discernable from that of tritium liberated into a stream of argon. Only the relative amount of the two released tritium-species, i.e. HT or HTO, was different. Temperature-dependent depth profiles could be predicted with a one-dimensional diffusion model. Diffusion coefficients derived from fitting of the tritium release into an evacuated vessel or a stream of argon were found to be (1.4 ± 1.0) × 10-7 and (1.3 ± 0.9) × 10-9 cm2/s at 573 and 423 K, respectively. Polished surfaces on type SS316 stainless steel inhibit considerably the thermal release rate of tritium.

Tritium Trapping on the Plasma Irradiated Tungsten Surface

Abstract Tungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.

Measurement of uptake and release of tritium by tungsten

Abstract The uptake of tritium by tungsten and its release behavior have been investigated. Specimens annealed at 773 K, 873 K, 973 K, 1,073 K and 1,173 K for 3 hours and loaded with tritium at 773 K for 3 hours accumulated 0.125 ppm, 0.068 ppm, 0.067 ppm, 0.038 ppm and 0.033 ppm, respectively (tritium solubilities were 3.6x 10-9 at.fr.Pa-1/2, 2.0x 10-9 at.fr.Pa-1/2, 1.9x 10-9 at.fr.Pa-1/2, 1.1x 10-9 at.fr.Pa-12/ and 9.7x 10-10 at.fr.Pa-1/2, respectively). The difference is attributed to the existence of trapping sites or oxide films.

Application of a Hydrothermal Treatment for the Decontamination from Tritium of Fusion Reactor Materials - Tritium Decontamination Using an Autoclave

Abstract A batch process concept for the decontamination from tritium of fusion reactor materials based on a hydrothermal treatment is under development at HRC. Essentially, tritium-loaded material is heated in a tightly closed vessel containing a defined amount of water. The objective of the water is to “capture” the released tritium in a small volume of liquid. For the detritiation, stainless steel temperatures in the range 393-473 K over a period of several days were found to be adequate. From the results it appears that by and large the released tritium accumulates in the purposely introduced water. The achieved degree of decontamination was estimated from the tritium concentration in the water and the tritium that remained in the decontaminated material. Tritium trapped in the surface layer of stainless steel was not reduced by the isochoric hydrothermal treatment in the same proportion as that in the bulk.