R.A. Langley

The beryllium limiter experiment in ISX-B

An experiment to test beryllium as a limiter material has been performed in the ISX-B tokamak. The effect of the plasma on the limiter and the effect of the limiter on the plasma were studied in detail. Heat and particle fluxes to the limiter were measured, and limiter damage by melting was documented as a function of power flux. Strong melting and evaporation of the limiter caused beryllium gettering of the vacuum vessel. Postmortem analysis of the limiter was performed to document the amount of retained hydrogen and the erosion and impurity deposition on the limiter. The effect of the limiter on the plasma performance was studied in terms of parameter space, impurity content, and confinement for the ungettered and gettered cases. Operational experience with beryllium in a fusion experiment is discussed.

Confinement improvement in beam heated ISX-B discharges with low-z impurity injection

Abstract Results are reported on improved confinement in the Impurity Study Experiment (ISX-B) neutral beam heated plasmas when a small amount of neon is injected shortly after the start of beam heating. The scaling of energy confinement is modified by the introduction of a dependence on line-averaged density. Calculations show the improvement is primarily caused by a reduction in electron heat conduction.

Hydrogen trapping, diffusion, and recombination in austenitic stainless steels

Trapping, diffusion, and recombination of hydrogen in austenitic stainless steels are reviewed. It is suggested that since all of these processes are strongly interdependent and since the measured recombination rates are found to vary four orders of magnitude at any temperature, the data analysis techniques used to date are insufficient. A two-region diffusion model with surface recombination is proposed in which the surface layer is characterized by a smaller diffusion coefficient than the bulk.

Plasma-wall impurity experiments in ISX-A☆

The ISX-A was a tokamak designed for studying plasma-wall interactions and plasma impurities. It fulfilled this role quite well, producing reliable and reproducible plasmas which had currents up to 175 kA and energy containment times up to 30 ms. With discharge precleaning, Zeff was as low as 1.6; with titanium evaporation, Zeff approached 1.0. Values of Zeff≳ 2.0 were found to be proportional to residual impurity gases in the vacuum system immediately following a discharge. However, there was no clear dependence of Zeff on base pressure. Stainless steel limiters were used in most of the ISX-A experiments. Upon introducing carbon limiters into the vacuum system, Zeff increased to 5.6. After twelve days of clean-up with tokamak discharges, during which time Zeff steadily decreased, the carbon limiters tended to give slightly higher values of Zeff than stainless steel limiters. Injection of <1016 atoms of tungsten into discharges caused the power incident on the wall to double and the electron temperature profile to become hollow.

Particle removal with pump limiters in ISX-B

Abstract The first pump limiter experiments were performed on ISX-B. Two pump limiter modules were installed in the top and bottom of one toroidal sector of the tokamak. The modules consist of inertia cooled, TiC-coated graphite heads and ZrAl getter pumps each with a pumping speed of 1000–2000 l/s. The objective of the initial experiments was the demonstration of plasma particle control with pump limiters. The first set of experiments were performed in ohmic discharges (OH) in which the effect of the pump limiters on the plasma density was clearly demonstrated. In discharges characterized by Ip = 110 kA, B T = 15 kG , n e = 1−5 × 10 13 cm −3 and t = 0.3 s, the pressure rise in the pump limiters was typically 2 mTorr with the pumps off and 0.7 mTorr after activating the pumps. When the pumps were activated, the line-average plasma density decreased by up to a factor 2 at identical gas flow rates. The second set of measurements were performed in neutral beam heated discharges (NBI) with injected powers between 0.6 MW and 1.0 MW. Due to a cooling problem on one of the ZrAl pumps, the NBI experiments were carried out with one limiter only. The maximum pressure observed in NBI-discharges was 5 mTorr without activating the pumps, i.e., approximately twice as high as in OH-discharges. The exhaust efficiency, which is defined as the removed particle flux divided by the total particle flux in the scrape-off layer, is estimated to be 5%.

Initial testing of TiB2 and TiC coated limiters in ISX-B

Abstract Low-Z coatings on graphite substrates have been developed for testing as limiters in the Impurity Study Experiment (ISX-B) tokamak. Laboratory and tokamak testings have been accomplished. The laboratory tests included thermal shock experiments by means of pulsed e-beam irradiation, arcing experiments, and hydrogen and xenon ion erosion experiments. The tokamak testing consisted of ohmically heated plasma exposures with energy depositions up to 10 kJ/discharge on the limiters. The coatings, applied by chemical vapor deposition, consisted of TiB 2 and TiC deposited on POCO graphite substrates. The limiter samples were interchanged through the use of a transfer chamber without atmospheric exposure of the ISX-B tokamak. Limiter samples were baked out in the transfer chamber before use in the tokamak. Provisions for both heating and cooling the limiter during tokamak discharges were made. Initial testing of the limiter samples consisted of exposure to only ohmically heated plasmas; subsequent testing will be performed in neutral-beam-heated plasmas having up to 3 MW of injected power. Bulk and surface temperatures of the samples were measured to allow the determination of energy deposition. Extensive plasma and edge diagnostics were used to evaluate the effect of the limiter on the plasma (e.g. vacuum ultraviolet spectrometry to determine plasma impurity concentrations, Thomson scattering to determine Z effective, IR camera to measure limiter surface temperature, and laser fluorescence spectrometry to determine neutral impurity concentration and velocity distribution in the limiter region).

Rotation scalings and momentum confinement in neutral-beam-injected ISX-B plasmas

Scalings of the central rotation in non-gettered, co-injected ISX-B discharges have been measured as a function of beam power, electron density and plasma current. Extensive studies are made possible by exploiting charge-exchange excitation (CXE) of 0 VIII lines to measure Doppler shifts. The rotation velocity, v(0), tends to saturate at (1.0 − 1.2) × l07 cms−1 when Pb0.5 MW, showing little further increase up to the maximum input of 2 MW; v(0) is independent of ne and Ip. Momentum confinement times in quasi-steady plasmas are 10–16 ms for e = 4.5 × 1013 cm−3. Counter-injection discharges always disrupt, but before this event v(0) is the same as for co-injection plasmas. The addition of a third beam line, permitting injection of up to 2 MW of balanced neutral-beam power, has allowed comparisons of the energy and particle confinement in rotating and non-rotating plasmas with the same total neutral-beam input. In those cases where impurity buildup can be avoided, it is found that the ISX-B empirical scaling of energy confinement time is reproduced with balanced injection. Thus, the unfavourable dependence of is not the result of rotation. Studies of impurity behaviour under differing injection conditions have been extended to include fully stripped low-Z ions. The results are consistent with previous investigations of metallic elements which revealed strong dependences on the sense (co versus counter) of injection. The potentials calculated from momentum balance, using measured rotation profiles and typical plasma density and temperature profiles, are in qualitative agreement with the potentials measured directly for various combinations of co- and counter-injection.

Radiative losses and improvement of plasma parameters after gettering in the advanced toroidal facility

The characteristics of plasmas in the Advanced Toroidal Facility (ATF) have proven to be strongly dependent on the type of wall conditioning employed. A succession of techniques, beginning with glow discharge cleaning and baking, and evolving to gettering with chromium and titanium, have led to progressive improvement of the plasma parameters. Gettering with titanium has reduced the low-Z impurity content by a factor of 3, lowered the radiated power by a factor of 2.5–3.5, and improved the control over the electron density. The maximum values achieved for stored energy, line averaged density and confinement times are 28 kJ, 1.2 × 1014cm−3 and 25 ms, respectively. These parameters are comparable to the best results achieved in the ISX-B tokamak which had the same average minor radius and one half the major radius of ATF. Quasi-steady operation for 200 ms of neutral beam injection (NBI) has been obtained in high density, titanium gettered plasmas without the collapses that were typical earlier periods of operation. Neon injection experiments have helped to delineate the limits on the global levels of radiation that can be maintained and have supported the conclusion that mechanisms other than radiative losses are important for initiating the collapses still observed in low density NBI plasmas.

Helium-ion-induced release of hydrogen from graphite☆

Abstract The ion-induced release of hydrogen from AXF-5Q graphite was studied for 300- to 500-eV helium ions. The hydrogen was implanted into the graphite with a low energy (~ 200 eV) and to a high fluence. This achieved a thin (~10 nm), saturated near-surface region. The release of hydrogen was measured as a function of helium fluence. A model that includes ion-induced detrapping, retrapping, and surface recombination was used to analyze the experimental data. The release cross section varied from 1.8 × 10 −16 cm 2 at 300 eV to 1.4 × 10 −16 cm 2 at 500 eV, while the depletion depth varied from 0.8 nm at 300 eV to 1.2 nm at 500 eV.

Vacuum pumping requirement considerations for future fusion devices

Abstract The vacuum pumping requirements for a fusion device are dictated by a number of factors, including the materials used in construction of the device, the cleaning and conditioning techniques implemented, the operating conditions and device temperature, the plasma characteristics, and the fuel gases and impurities retained and released by the plasma-facing components (PFCs). In an attempt to derive guidelines for determining the vacuum pumping requirements of a generic fusion device, a study was undertaken to examine the vacuum pumping capabilities of existing large fusion devices, to review the cleaning and conditioning techniques now in use, and to catalog pertinent vacuum equipment that is now available or anticipated soon. In a survey of six large fusion devices [ASDEX, DIII-D, the Joint European Torus (JET), JT-60, the Tokamak Fusion Test Reactor (TFTR), Tore Supra], information was collected on PFC materials, cleaning and conditioning techniques, device operating temperatures, and vacuum pumping system characteristics and on the surface conditions and partial pressures necessary to achieve high-purity plasma discharges in each device. The results of the survey are reviewed to determine how the various factors affect vacuum pumping requirements. The use of tritium in future fusion devices is expected to create a new set of problems for the vacuum pumping systems of these devices. These problems are addressed, and some possible solutions for the pumping and gas handling systems are identified. New vacuum equipment is under development, and the fusion community is in a position to influence its direction and emphasis. Information gained in studies of this type should be brought to bear on the development process to ensure that vacuum equipment will be available to meet the needs of future fusion devices.