K.L. Wilson

Deuterium trapping in irradiated 316 stainless steel

Abstract Linear ramp thermal desorption measurements were conducted on 316 stainless steel samples implanted at 296 K with 1–10 keV D + to fluences of 10 17 –10 19 D + /cm 2 . Samples were held 1–100 h at 296 K prior to desorbing. The desorption data were shown to arise from two dominant mechanisms: bulk migration of mobile deuterium atoms with ~0.6 eV migration energy, and release from near surface traps with a net detrapping energy of ~0.9 eV. After 10 keV D + bombardment, more complex desorption spectra were observed. Samples pre-damaged with 300 keV He + exhibited a significant increase in the deuterium trapping compared to samples without pre-bombardment. Based on the data and modelling, estimates of tritium retention in TFTR were made.

Tritium retention in tungsten exposed to intense fluxes of 100 eV tritons

Abstract Tungsten is a candidate material for the International Thermonuclear Experimental Reactor (ITER) as well as other future magnetic fusion energy devices. Tungsten is well suited for certain fusion applications in that it has a high threshold for sputtering as well as a very high melting point. As with all materials to be used on the inside of a tokamak or similar device, there is a need to know the behavior of hydrogen isotopes embedded in the material. With this need in mind, the Tritium Plasma Experiment (TPE) has been used to examine the retention of tritium in tungsten exposed to very high fluxes of 100 eV tritons. Both tungsten and tungsten containing 1% lanthanum oxide were used in these experiments. Measurements were performed over the temperature range of 423–973 K. After exposure to the tritium plasma, the samples were transferred to an outgassing system containing an ionization chamber for detection of the released tritium. The samples were outgassed using linear ramps from room temperature up to 1473 K. Unlike most other materials exposed to energetic tritium, the tritium retention in tungsten reaches a maximum at intermediate temperatures with low retention at both high and low temperatures. For the very high triton fluences used (>10 25 T/m 2 ), the fractional retention of the tritium was below 0.02% of the incident particles. This report presents not only the results of the tritium retention, but also includes the modeling of the results and the implication for ITER and other future fusion devices where tungsten is used.

Hydrogen recycling properties of stainless steels

Abstract The hydrogen retention and release characteristics of stainless steels that contribute to plasma-wall recycling of a magnetically confined plasma are reviewed. Details are presented on laboratory measurements of hydrogen reflection, desorption, trapping and release. Critical data needs are shown to include reflection below 100 eV, desorption cross sections at realistic energies, hydrogen surface coverage, and molecular recombination rates for characterized first wall surfaces.

Hydrogen recycling properties of graphite

Abstract A review is presented on the hydrogen recycle properties of graphite, the leading candidate material for plasma interactive component surfaces. The hydrogen transport database and modelling are summarized, followed by a discussion of its recycling properties in operating devices and its predicted tritium retention behavior in future D-T reactors.

Bulk-boronized graphites for plasma-facing components in ITER

Abstract Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt% to 30 wt% have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600 °C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2–3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt% bulk-boronization at temperatures above 1000 °C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt% bulk-boronization of graphite hinders air oxidation nearly completely at 800° C and reduces the steam oxidation rate by a factor of 2–3 at around 1100 and 1350 °C.

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

Trapping of deuterium at damage in graphite

Enhanced retention of deuterium (D) and tritium (T) in graphite due to neutron damage could result in large T inventories in future DT fueled fusion devices such as CIT and ITER. This paper describes experiments done to characterize the effect of lattice damage on retention of D in graphite. Lattice damage was produced in several types of graphite by irradiation with 6 MeV C+ ions or with neutrons. The damaged graphite was then heated to 1200{degree}C and exposed to D{sub 2} gas. The concentration of D retained in the graphite was measured by nuclear reaction analysis and compared to the level of damage. In POCO, N3M and H451 graphites, D retention increased with damage at low damage levels but saturated above 0.04 dpa at a D concentration of about 650 atomic ppM. D retention in damaged highly oriented pyrolytic graphite was an order of magnitude smaller than in the other graphites indicating that crystalline microstructure is an important factor. These results indicate that neutron damage in CIT and ITER may cause retention of large inventories of tritium in graphite components. 5 refs., 4 figs., 1 tab.

Low-energy helium implantation of aluminum

A series of 20 keV He/sup +/ implantations was conducted on well-annealed MARZ grade aluminum at fluxes of 6 x 10/sup 14/ and 6 x 10/sup 13/ He/sup +//cm/sup 2/ sec. Three distinct, temperature dependent He release mechanisms were found by He re-emission measurements during implantation, and by subsequent SEM and TEM investigations. At 0.08 of the melting temperature (T/sub m/) gas re-emission rose smoothly after a critical dose of 3 x 10/sup 17/ He/sup +//cm/sup 2/, with extensive blistering. The intermediate temperature range (approximately 0.3 T/sub m/) was characterized by repeated flake exfoliation and bursts of He after a dose of 3 x 10/sup 17/ He/sup +//cm/sup 2/. Rapid He evolution, with hole formation was found above 0.7 T/sub m/. No significant differences in either gas re-emission or surface deformation were found between the two fluxes employed.

Retention of deuterium and tritium in Papyex graphite

Abstract The retention of 100 eV D-T ions in Papyex graphite over the temperature range 373 to 1573 K has been measured using both nuclear reaction analysis and tritium dissolution counting. Below 1000 K, the retention is characterized by saturation of the near surface region and atomic adsorption on internal porosity, both of which decrease with increasing temperature. Above 1000 K, a local maximum in the retention is seen near 1200 K due to intergranular diffusion and decoration of high energy traps.

Deuterium trapping and release in titanium-based coatings for TFTR☆

Abstract We have measured the deuterium trapping and thermal release characteristics of three candidate armor and limiter materials for TFTR: Ti explosively bonded onto Cu; TiB 2 and TiC chemically vapor deposited on C. The re-emission rate of deuterium was monitored during bombardment with 10 keV D + 3 and the post-implantation retention was measured with D( 3 He, α)H nuclear reaction profiling and linear ramp thermal desorption. Titanium claddings showed no thermal release of deuterium for fluences of 10 19 D/cm 2 at 375 K; at 775 K, only 65% of the fluence was released during implantation to 10 18 D/cm 2 . This high retention is shown to result from titanium deutende precipitation and the formation of a bulk deuterium solid solution. In contrast, the TiB 2 and TiC coatings showed a rapid saturation in deuterium retention of 17 D/cm 2 .