Publications

Efficient techniques for rapid tritium removal will be necessary for ITER to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO{sub 2} or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm{sup 2} flux with an exposure time of order 10 ms is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally.

Tritium retention and removal are critical issues for the success of ITER or any DT fusion reactor. The Tokamak Fusion Test Reactor, TFTR, is the first fusion facility to afford the opportunity to study the tritium retention and removal over an extended period. In TFTR, tritium accumulates on all surfaces with line of sight to the plasma by codeposition of tritium with carbon. Measurements of both deuterium and tritium retention fractions have yielded retention between 0.2 and 0.6 of the injected fuel in the torus. Tritium has been successfully removed from TFTR by glow discharge cleaning and by air purges. The in-vessel inventory was reduced by a factor of 2, facilitating machine maintenance. In TFTR, the amount of dust recovered from the TFTR vacuum vessel has varied from several grams to a few kilograms.

In applications dealing with the deposition of amorphous hydrogenated carbon layers or in the determination of the composition of deposited layers on the walls of nuclear fusion plasma experiments, the analysis of mixtures of light elements on heavy substrates is necessary. Depth profiling by means of RBS is often difficult due to the overlap of the backscattering intensities of different constituents from different depths. The erosion and reaction of deposited amorphous deuterated carbon (a-C:D) films with a Be substrate due to annealing in air poses an analytical challenge especially if simultaneously the exchange of hydrogen isotopes should be monitored. The analysis of the different recoiling atoms from collisions with heavy ions in Elastic Recoil Detection (ERD) can provide a tool which resolves all constituents in a single analysis. In the present study the composition of intermixed layers on Be containing H, D, Be, C and O has been analyzed using conventional {sup 4}He RBS at 2.2 MeV together with 2.5 MeV {sup 4}He ERD for hydrogen isotope analysis. At these energies, an overlap of signals from different constituents could be avoided in most cases. As alternative method heavy ion ERD using Si{sup 7+} ions extracted from a 5 MeV Tandem Van de Graff accelerator was investigated. At a scattering angle of 30{degree} Si ions could not be scattered into the detector and a solid state detector without protecting foil could be used. Even in the intermixed layers at terminal energies of 5 MeV the heavy constituents could be separated while signals from recoiling hydrogen and deuterium atoms could be resolved on top of the signal from the Be substrate. For the analysis of the RBS and ERD data the newly developed spectra simulation program SIMNRA has been used which includes a large data bank for scattering and nuclear reaction cross sections. The depth profiles of all constituents extracted from the simulation are compared for both methods.

Thin metal films (∼ 100 nm thick) of Be, W, V and Mo, were deposited on a Si depth-marked graphite sample and exposed to the steady-state outer strike point on DIII-D in order to measure their respective erosion rates. Gross erosion rates and redeposition lengths are found to decrease with the atomic number of the metallic species, as expected. The maximum net erosion rate for carbon, which occurs near the separatrix, increased from 4 to 16 nm/s when the incident heat flux was increased from 0.7 to 2 MW/m2. Comparisons of the measured carbon erosion with REDEP code calculations show good agreement for both the absolute net erosion rate and its spatial variation. Visible spectroscopic measurements of singly ionized Be (BeII 4674 Å) have determined that the erosion process reaches steady-state during the exposure.

Measurements of the deuterium particle flux and energy to the divertor of the DIII-D tokamak during a series of plasmas that terminated in disruptions have been made using a silicon collector probe installed on the DiMES (divertor materials exposure system) mechanism. During the steady state portion of each discharge, the probe was located in the private flux region, but immediately before disrupting the plasma, by injecting either Ar or D2 gas, the strike point of the outer divertor leg was positioned over the probe. Comparison of the amount of retained D in the probe for the two types of disruptions indicates that much of the trapped D could have resulted from exposure in the private flux zone prior to the disruption. Measurements of the depth distribution of the trapped D in the Si imply that the incident ion energy was approximately 100 eV at normal incidence and decreased slightly at oblique angles. The measurements give an upper bound to the energy of deuterons striking the divertor floor in the vicinity of the strikepoint during disruptions.

Lithium wall conditioning has been used in a recent campaign evaluating high performance negative central shear (NCS) discharges. During this campaign, the highest values of stored energy (4.4 MJ), neutron rate (2.4 x 10{sup 16}/s), and nT{sub i}{tau} (7 x 10{sup 20} m{sup -3}-keV-s) achieved to date in DIII-D were obtained. High performance NCS discharges were achieved prior to beginning lithium conditioning, but it is clear that shot reproducibility and performance were improved by lithium conditioning. Central and edge oxygen concentrations were reduced after lithium conditioning, Lithium conditioning, consisting of up to four pellets injected at the end of the preceding discharge, allowed the duration of the usual inter-shot helium glow discharge to be reduced and reproducible high auxiliary power discharges, P{sub NBI} {<=} 22 MW, were obtained with plasma currents up to 2.4 MA.

Plasma disruptions are a serious concern in tokamak design because of the high impulsive heat loads which can cause strong erosion of divertor materials due to enhanced sputtering, or melting/ablation in the most severe cases. Predictions of net erosion rates and hence component lifetimes are very difficult and are highly dependent on the plasma conditions over the divertor target. It is therefore necessary to characterize the properties of the scrape-off plasma near the divertor target plate under these special conditions. Here, plasma/wall interaction studies are being carried out using the Divertor Materials Exposure System (DiMES) on DIII-D. The objective of the experiment is to determine the kinetic energy and flux of deuterium ions reaching the divertor target during argon-induced radiative disruptions. The experiment utilizes a special slotted ion analyzer mounted over a Si sample to collect the fast charge-exchange (CX) deuterium neutrals emitted within the recycled cold neutral layer (CNL) which serves as a CX target for the incident ions. A theoretical interpretation of the experiment reveals a strong forward pitch-angle dependence in the approaching ion distribution function. The depth distribution of the trapped D in the Si sample was measured using low-energy direct recoil spectroscopy. Comparison with the TRIM code using monoenergetic ions indicated that the best fit to the data was obtained for an ion energy of 100 eV. An estimate of the CNL thickness {integral}nd{ell} indicates that during disruptions the CNL cushion is thick enough to reduce the local ion heat load by {approximately}30% due to CX refluxing.

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 a 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 the water present in 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. 24 refs., 2 figs.