Y. Hirooka

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.

Performance of boron/carbon first wall materials under fusion relevant conditions

Abstract The conditioning of the plasma facing wall in thermonuclear confinement experiments has been performed very successfully by the application of amorphous boron containing hydrogenated carbon films. Boronization leads to tokamak discharges with significantly reduced oxygen and carbon contaminations. For high heat flux components (especially in future quasi-stationary confinement experiments) new boron/carbon materials have to be developed: monolithic tiles of boronated graphites which can be brazed to watercooled substrates or thick B 4 C-coatings on graphite or high-Z coolant tubes. A variety of bulk materials (boronated graphites with boron contents in the range from 3 to 30%, so-called coat mix material on the basis of B 4 C) and coatings (amorphous B/C films, thick B 4 C layers applied by LPPS or CVD methods) were characterized systematically. In addition the behaviour of these materials was investigated under thermal loads; erosion and disruption simulation experiments were performed in electron and ion beam high heat flux test facilities. Physical and chemical sputtering of the coat-mix-material was studied in the PISCES-B facility in dependence on the hydrogen ions fluence.

Evaluation of tungsten as a plasma-facing material for steady state magnetic fusion devices

Tungsten in the form of bulk-material, and relatively thick (1 mm) chemically deposited and plasma-sprayed coatings on molybdenum, has been evaluated as a plasma-facing material for near future steady state magnetic fusion devices, focusing on issues related to the divertor plate design. These issues are: (1) thermal outgassing; (2) plasma erosion; (3) deuterium retention; (4) disruption erosion; and (5) surface modifications. Total outgassing quantities from bulk tungsten and chemically deposited coatings are substantially smaller than those from graphites. Effects of redeposition and impurities on the erosion behavior due to deuterium plasma bombardment have been analyzed. Trace amounts of oxygen-containing impurities in the plasma can reduce the threshold energy for physical sputtering, affecting the overall erosion behavior of tungsten at energies below 500 eV. However, it has been found that at electron temperatures around 5 eV or lower, fragmentation of these impurities followed by positive ionization is significantly reduced, whereby plasma erosion data basically agree with sputtering theories and ion beam data. Thermal desorption measurements after plasma bombardment have indicated that the deuterium retention quantity in tungsten materials is of the order of 10 14–15 D atoms/cm 2 . At simulated disruption with an energy deposition of 6 MJ/m 2 , the effective heat deposition is found to be reduced to about 1%, due to a combined effect of molten layer protection and vapor shielding. Steady state plasma bombardment removes surface impurities and smooths the surface topography along with surface erosion whereas disruption causes microscopic cracking and surface melting.

Erosion and redeposition experiments in the PISCES facility

The modification of surfaces during exposure to plasma bombardment is a critical issue in the development of limiter and wall materials for fusion confinement experiments. Controlled studies of the erosion and redeposition of materials during high flux and fluence plasma exposure are now possible in the PISCES facility. PISCES is a continuously operating plasma device which has achieved hydrogen plasma densities of over 1013 cm−3 and electron temperatures of 5 to 24 eV over large areas. Ion fluxes of 1017 to 1019 cm−2 s−1 and fluences of up to 1023 cm−2 have been used to bombard biased samples inserted into th plasma. The plasma parameters can be selected to produce simple sputtering, or redeposition by the ionization and recycling of the sputtered target materials. Collaborative studies on the performance of Cu and Cu-Li alloys (with ANL), stainless steel (with SNLL), and graphite (with IPP at Garching, and SNLL) have been undertaken. Surface topography modification is always observed after a sufficient fluence is achieved. The net erosion rate is significantly lower during redeposition than one would expect from classical sputtering yields. The transport and deposition of different materials by the plasma to the samples during redeposition conditions results in greatly modified surface composition and morphology. Chemical sputtering of graphite during low energy, high flux (>1018 cm−2 s−1) plasma bombardment is observed. Chemically formed hydrocarbons are relatively easily redeposited compared to sputtered carbon. The performance of these materials, the surface morphology evolution, and the characteristics of the redeposited materials are discussed.

Presheath profiles in simulated tokamak edge plasmas

Abstract Steady state magnetized plasmas produced by the PISCES experiment are used to study plasma-wall interaction phenomena relevant to confinement devices such as tokamaks. An experimental investigation of the presheath region that extends from a wall surface into “simulated tokamak” edge plasmas along magnetic field lines is reported. Diagnostics especially developed for this work include a fast-scanning multiple Langmuir/Emissive/Mach probe system and a CID camera imaging system. Measurements of density, electron temperature, floating potential, space potential, and bulk plasma flow velocities have been obtained in plasmas with densities ranging from 1012 to 1013 cm−3, electron temperatures from 5 to 15 eV. and axial magnetic fields of 0.2 to 1.4 kG. Plasma density profiles along the magnetic field typically show a characteristic factor of 2 decrease towards the wall surface. A plasma potential variation in the near presheath zone of order 0.5 T e is measured, consistent with the bulk plasma flow approaching the ion sound speed near the wall surface, as inferred from a simple “free fall” model. A Boltzmann model for the presheath density profile accuracy tracks the density profile measured both by the Mach probe and by spectroscopic means. Flow profiles are used as a consistency check on various magnetized Mach probe theories. Results suggest that cross-field transport of parallel momentum through viscosity is relatively unimportant in PISCES plasmas and thus may be unimportant in tokamak boundary layer plasmas. Discharges with non-thermal electrons display axial profiles of space potential and floating potential which indicate a “hotter” electron distribution function near the wall surface, consistent with “colder” electrons being reflected by the presheath potential drop.

Impurity transport and retention in a gas target divertor: simulation experiments in PISCES-A and modeling results

Impurity retention in the gaseous divertor regime is investigated in the PISCES-A facility at UCLA. We report measurements and 1 1/2D fluid modeling results of impurity transport for typical tokamak divertor plasma parameters (10 18 ≤ n e ≤3×10 19 m −3 , kT e ≤20 eV). The neutral hydrogen density close to the (simulated) divertor target is 10 20 ≤ n 0 ≤3×10 21 m −3 . Gaseous trace impurities (argon, neon) as well as low- Z and high- Z materials sputtering carbon, tungsten) are studied. It is observed that the impurity retention in a gaseous divertor is substantially improved as compared to conventional divertor operating regimes. The modeling results suggest that the retention of neutral and ionized impurities is mainly due to collisions with hydrogen (deuterium) neutrals and ions streaming towards the divertor target a a velocity of 0.25–0.5 c s . A low level of residual impurity transport, observed at high neutral density, is attributed to a plasma flow reversal close to the radial boundary. Sputtering of a tungsten sample by intrinsic impurities has been shown to decrease substantially for target electron temperatures kT e

Helium line emission measurements in PISCES-B as a tool for Te-profile determinations in tokamak boundary plasmas

Abstract The potential of helium spectroscopy for quasicontinuous electron temperature profile measurements in the boundary layer of a tokamak plasma is tested. Particularly thermal atomic helium beams are easy to produce and can penetrate relatively far into typical tokamak boundary plasmas because of their high ionization energy. However, in the case of helium and for the range of densities and temperatures characteristic for tokamak boundaries the validity of the corona model cannot be assumed. Therefore it is vital to measure Hel-line intensities in well diagnosed plasmas with tokamak boundary like conditions, which may be compared with calculated ones. For the production of a plasma with parameters close to that of a tokamak boundary layer the PISCES-B facility was used, where plasmas with up to T e ≈ 30 eV simultaneously with n e ≈ 5 × 10 12 /cm 3 could be obtained. T e - and n e -profiles were determined both by fixed and reciprocating Langmuir probes. The helium was injected into the central part of an argon plasma, and its spectral line emission in the visible was measured by means of an OMA attached to an optical spectrometer. Six triplet and six singlet emission lines of HeI were measured and the ratio of these — which is assumed to be a function of T e — evaluated for a range of temperatures and densities. From their absolute intensities also the individual line excitation rates could be derived as a function of electron temperature. It was found that the majority of the experimental values both for the line excitation rate and the line intensity ratio do not agree with the calculated values, which take only excitation from the ground state into account. On the other hand, experimental intensity ratios were found, which do not show a density dependence, and are recommended to be used for an experimental technique, which would allow a quasicontinuous T e -profile measurement in the plasma boundary region of tokamaks.

Comparison of three boronization techniques in TdeV

Preparation of the internal walls of tokamaks by plasma enhanced chemical vapour deposition (PECVD) of boron containing films has now been implemented on several machines since its development on TEXTOR. More recently, such films were deposited on the internal walls of TdeV using not only this procedure but also two new approaches: solid target boronization (STB) which consisted in inserting a low-density boronized carbon-carbon (C-C) composite into the tokamak plasma and TMB fuelling where trimethylboron was used as fuelling gas during the plasma discharge. These approaches resulted in a rapid shot to shot improvement of important parameters such as the volume averaged resistivity and radiated power over the first dozen shots when the boron source is present. Typically, the resistivity is reduced from ∼4.0×10 −7 to ∼2.5×10 −7 Ωm, comparable to the resistivity obtained with PECVD. The radiated power relative to the ohmic power is reduced by a factor of 2 from 20 to 10%. When the boron source, present during STB or TMB fuelling, is removed however, these plasma parameters start increasing. Within a few tens of shots, they have reverted to their preconditioning values, a situation which requires hundreds of shots after PECVD.

Solid target boronization in the Tokamak de Varennes: a technique for real-time boronization

A new boronization technique, called solid target boronization (STB), has been developed and successfully applied in the Tokamak de Varennes. For this technique, a boronized carbon-carbon composite material is used as a source of boron and is exposed directly to confinement plasmas in order to induce boron emission due to thermal evaporation and sputtering. The STB technique can provide real-time boronization effects by implementing self-reconstructing boron-carbon coatings. A significant decrease in impurity radiation due to oxygen as well as several metals has been observed. Also, the fuelling rate increases considerably during STB, which is attributed to codeposition of carbon, boron and hydrogen

The TEXTOR helium self-pumping experiment: Design, plans, and supporting ion-beam data on helium retention in nickel

Abstract A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, ~ 25 × 25 cm2 heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of ~ 50Anickel will be deposited on the two trapping plates. During a shot helium and hydrogen ions will impinge on the plates through a ~ 3 cm wide entrance slot. The helium removal capability, due to trapping in the nickel, will be assessed for a variety of plasma conditions. In support of the tokamak experiment, the trapping of helium over a range of ion fluences and surface temperatures, and detrapping during subsequent exposure to hydrogen, were measured in ion beam experiments using evaporated nickel surfaces similar to that expected in TEXTOR. Also, the retention of H and He after exposure of a nickel surface to mixed He/H plasmas has been measured. The results appear favorable, showing high helium trapping (~ 10–50% He/Ni) and little or no detrapping by hydrogen. The TEXTOR experiment is planned to begin in 1991.