R.A. Causey

Hydrogen isotope retention and recycling in fusion reactor plasma-facing components

The proper design of a fusion reactor is not possible unless there is an understanding of the hydrogen isotope retention and recycling for the plasma-facing components. From the tritium inventory point of view, it is absolutely necessary to understand the short-term and long-term hydrogen isotopes retention characteristics of the individual materials used for the first wall and divertor. From the plasma density and fueling point of view, it is necessary to understand the recycling characteristics of these materials. This report is an overview of the available data on hydrogen isotope retention and recycling for beryllium, tungsten, carbon, and selected liquid metals. For each material discussed, recommendations are made as to the most appropriate values to use for parameters such as diffusivity, solubility, recombination rate coefficient, and trapping.

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.

The interaction of tritium with graphite and its impact on tokamak operations

Abstract The state of knowledge of tritium retention in graphite and how it affects the operation of tokamak fusion reactors is reviewed. The retention of tritium in the saturated surface layer exposed to the plasma, retention in the surface connected porosity, retention in the graphite grains, and retention in the co-deposited carbon/tritium layer are discussed along with an assessment of the relative importance of each component for fusion reactors. Data on the diffusivity and solubility of tritium in graphite are analyzed to pick the “best estimate” for both of these parameters. Tritium trapping in graphite is considered along with an explanation for its cause and an estimate of its magnitude.

Behavior of tungsten exposed to high fluences of low energy hydrogen isotopes

Abstract Tungsten is a candidate plasma facing material under investigation in a Sandia National Laboratories project conducted at Los Alamos National Laboratory. Samples of 99.95% tungsten provided by Plansee Aktiengesellschaft were exposed to 100 eV deuterium and tritium ions at a range of fluxes from 2.3×1017 to 1.3×10 18 ions / cm 2 s for one hour at 623 K in the tritium plasma experiment. The samples were outgassed to determine the amount of retained hydrogen isotopes. The retention scaled at slightly greater than the square root of the fluence. The fractional retention was on the order of 10−5. The data from these experiments were combined with previous results to construct a comprehensive model of the migration and retention behavior for hydrogen in tungsten. A second set of experiments involved exposing 99.95% tungsten foils provided by AESAR to 100 eV deuterons at a flux of 6×10 17 D / cm 2 s for 30 min at 423 and 373 K. Scanning Electron Microscopy analysis was performed on the samples to determine the effects of the plasma exposure. Unannealed samples revealed extensive blistering with many blister caps removed. Samples annealed to 1473 K showed minor blistering, and samples annealed to 1273 K showed no blistering. The SEM analysis was used in conjunction with the retention results to understand the role of annealing and defects in trapping within the tungsten.

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.

Tritium Inventory in the ITER PFC's: Predictions, Uncertainties, R&D Status and Priority Needs

Abstract New data on hydrogen plasma isotopes retention in beryllium and tungsten are now becoming available from various laboratories for conditions similar to those expected in the International Thermonuclear Experimental Reactor (ITER) where previous data were either missing or largely scattered. Together with a significant advancement in understanding, they have warranted a revisitation of the previous estimates of tritium inventory in ITER, with beryllium as the plasma facing material for the first-wall components, and tungsten in the divertor with some carbon-fibre-composites clad areas, near the strike points. Based on these analyses, it is shown that the area of primary concern, with respect to tritium inventory, remains codeposition with carbon and possibly beryllium on the divertor surfaces. Here, modelling of ITER divertor conditions continues to show potentially large codeposition rates which are confirmed by tokamak findings. Contrary to the tritium residing deep in the bulk of materials, this surface tritium represents a safety hazard as it can be easily mobilised in the event of an accident. It could, however, be possibly removed and recovered. It is concluded that active and efficient methods to remove the codeposited layers are needed in ITER and periodic conditioning/cleaning would be required to control the tritium inventory and avoid exhausting the available fuel supply. Some methods which could possibly be used for in-situ cleaning are briefly discussed in conjunction with the research and development work required to extrapolate their applicability to ITER.

Comparison of the thermal stability of the codeposited carbon/hydrogen layer to that of the saturated implant layer

Abstract 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.

The effect of surface composition on plasma driven permeation of deuterium through 304 stainless steel

Abstract The effect of incident ion energy on plasma driven permeation of deuterium through 304 stainless steel has been studied experimentally. The relationship of permeation rate to ion induced surface composition changes was determined by in situ Auger electron spectroscopy. The presence of either C or O on the upstream surface increased the permeation rate as compared to a sputter cleaned surface. C in the form of near surface carbide, however, had little or no effect on the permeation rate.

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.