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Tungsten is under consideration for use as a plasma-facing material in the divertor region of ITER. Lithiation can significantly improve plasma performance in long-pulse tokamaks like EAST. The investigation of lithiated tungsten is important for understanding the lithium conditioning effects for EAST, where tungsten will be used as a plasma-facing material. In this paper, a few important issues of lithiated tungsten interacting with high-flux deuterium plasma have been studied, such as the effect of lithiation on deuterium retention, the profile of elemental distribution, and the chemical state of lithiated tungsten. Deuterium retention inside both pure and lithiated tungsten has been investigated for the first time in the linear plasma simulator Magnum-PSI by in-situ laser induced breakdown spectroscopy (LIBS). The results indicate that, after deuterium plasma exposure, deuterium retention could be saturated in the lithiation layer, and the lithium in the lithiated layer is chemically bound with deuterium. Moreover, the lithiation can inhibit the blistering on the tungsten surface. These results can be valuable for the application of LIBS as a diagnostic technique for plasma-facing components of tokamaks.

The performance of the full-W ITER divertor may be significantly affected by the interplay between steady-state plasma exposure and transient events. To address this issue, the effect of a high-flux H plasma on the thermal shock response of W to ELM-like transients has been investigated. Transient heating of W targets is performed by means of a high-power Nd:YAG laser with simultaneous exposure to H plasma in the linear device Magnum-PSI. The effects of simultaneous exposure to laser and plasma have been compared to those sequentially and to laser only. Transient melting is found to be aggravated during plasma exposure and to occur at lower heat flux parameters. Roughness and grain growth are observed to be driven by peak temperature, rather than by the loading conditions. The temperature evolution of the W surface under a series of transients is recorded by fast infrared thermography. By accounting for changes in the reflectivity at the damaged surface as measured by ellipsometry, a reduction in power handling capabilities of the laser/plasma affected W is concluded. The evidence of reduced power handling of the W surface under conditions as described here is of great concern with respect to the durability of W PFCs for application in fusion devices.

The rate at which Li films will erode under plasma bombardment in the NSTX-U divertor is currently unknown. It is important to characterize this erosion rate so that the coatings can be replenished before they are completely depleted. An empirical formula for the Li erosion rate as a function of deuterium ion flux, incident ion energy, and Li temperature was developed based on existing theoretical and experimental work. These predictions were tested on the Magnum-PSI linear plasma device capable of ion fluxes >1024 m−2 s−1, ion energies of 20 eV and Li temperatures >800 °C. Li-coated graphite and TZM molybdenum samples were exposed to a series of plasma pulses during which neutral Li radiation was measured with a fast camera. The total Li erosion rate was inferred from measurements of Li-I emission. The measured erosion rates are significantly lower than the predictions of the empirical formula. Strong evidence of fast Li diffusion into graphite substrates was also observed.

During the C28-C30 campaigns the ITER-like Wall (ILW) was periodically inspected by IVIS (InVessel Inspection System) without breaking the vacuum. After these campaigns the JET chamber was opened and the wall was inspected with a high resolution camera. The marker tiles were removed from the chamber and they have been inspected in the beryllium handling facility. The overall picture of the ILW after more than 3000 pulses looks good. However there are a few modifications such as deposition of Be on some W coated tiles, local destruction of the W coating by runway electrons in an emergency shutdown, arcing on particular W coated tiles, small delaminations on G6 and G8 divertor tiles. These aspects together with the results of high heat flux tests carried out with W coated CFC samples in Magnum-PSI are discussed in the paper.

The plasma-surface interactions expected in the divertor of a future fusion reactor are characterized by extreme heat and particle fluxes interacting with the plasma-facing surfaces. Powerful linear plasma generators are used to reproduce the expected plasma conditions and allow plasma-surface interactions studies under those very harsh conditions. While the ion energies on the divertor surfaces of a fusion device are comparable to those used in various plasma-assited deposition and etching techniques, the ion (and energy) fluxes are up to four orders of magnitude higher. This large upscale in particle flux maintains the surface under highly non-equilibrium conditions and bring new effects to light, some of which will be described in this paper.

Previous studies have shown that transient thermal shock loads induce crack networks on tungsten samples especially at low base temperatures. To achieve test conditions which are more relevant for the performance of tungsten-armoured plasma facing components in next step thermonuclear fusion devices tungsten tiles were exposed to high flux hydrogen-plasma in the linear plasma generator Pilot-PSI and the high heat flux ion beam test facility MARION. Subsequently, the cyclic transient heat load tests were done in the electron beam facility JUDITH 1. The induced damages after these combined tests were examined by microscopically means, profilometry and metallography. The comparison of the obtained results and damage characteristics with those obtained after thermal shock loading show that the preloading of tungsten targets with high flux hydrogen-plasma has significant influence on the thermal shock behaviour of tungsten in terms of crack distance, width, and depth as well as cracked area. Furthermore the plasma parameters, in particular pulse duration and sample temperature during loading, have strong impact on the damage pattern after thermal shock loading. © 2013 Elsevier B.V. All rights reserved.

Helium (He) concentrations in tungsten nano-tendrils (W fuzz) have been measured for the first time using Elastic Recoil Detection (ERD). Fuzzy and non-fuzzy W surfaces were analyzed in order to illuminate the role of He in the transition in surface morphologies. Samples grown in the PISCES-A and PILOT-PSI experiments allowed a survey of surface temperature ranging from T s = 470–2595 K and of He fluence on the order of Φ He ∼ 10 24 –10 27 ions/m 2 . He concentrations measured in the bulk of W fuzz layers are roughly uniform with bulk He concentration 1–4 at.% while samples with just He in the near surface peaked at 1–2 at.%. This confirms that the nano-tendrils are filled with high pressure He bubbles since the solubility of He in W is ∼10 −5 at.%. This indicates that the ∼1000 K temperature fuzz-growth threshold is determined by the response of the W, not the near-surface He concentration.

We report measurements of chemical sputtering yields of graphite exposed to low temperature nitrogen plasmas. The influence of surface temperature and incoming ion energy on the sputtering yields has been investigated in two distinct ion flux density regimes. Sputtering yields grow consistently with increasing temperatures in experiments with low flux density (Gamma(i) approximate to 10(20) m(-2)s(-1) - 10(21) m(-2)s(-1)) and high flux density (Gamma(i) approximate to 10(23) m(-2)s(-1)). Moreover, empirical fitting of the data suggests that the temperature of 670 degrees C is optimal for chemical sputtering at high flux density. Negative biasing of the samples was used to vary the ion energy in the low flux density regime. The sputtering yield in this case increases from 0.07 atoms/ion for E-i = 1.5eV to 0.19 atoms/ion for E-i = 35 eV. After taking into account the dependence of the yields on temperature and ion energy, we evidenced a flux dependence of sputtering, similar to that found for chemical sputtering of carbon by hydrogen.

Experiments were performed in the Pilot-PSI linear plasma device to study the formation of helium-induced nanostructures under high heat fluxes (>10 MW m−2), the erosion of a tungsten surface at the elevated temperatures where those structures typically form and most importantly the behavior of a helium-induced fibreform structure during ELM-like pulses. A strong correlation between the characteristic size of the helium-induced morphology and the size of the voids observed in the near-surface region has been found, and both increase in size with increasing temperatures. Temperature-dependent erosion of tungsten surfaces was observed despite the ion energy being below the sputtering threshold, with an exponential increase of the erosion for temperatures higher than 2400 °C. The same effect was observed during ELM-like plasma pulses. Finally, a complete disappearance of the tungsten fibreform structure was observed after only one plasma pulse for energy densities higher than 0.5 MJ m−2.

Accidental melting of metallic plasma-facing materials in future fusion devices poses serious issues regarding the material lifetime and power-handling capabilities as well as core plasma performances. The behaviour of aluminium (as a proxy for beryllium) and tungsten materials was investigated in the Pilot-PSI linear plasma device to study the melt-layer motion and droplet ejection under ITER-relevant plasma conditions. Heat fluxes of up to 50 MW m(-2) raised the surface temperature to values up to 5000 K. The melt-layer rotation was found to depend on the magnetic field (up to 1.6 T) strength and target potential and is attributed to J x B forces caused by radial currents in the plasma. The amount of droplets ejected from the molten surface depends on the material-more droplets ejected from aluminium than from tungsten-and the heat flux to the target. The average droplet velocity was determined to be around 60 m s(-1) for both materials with droplets being ejected mainly in the axial direction. Droplet ejection is only observed during helium discharges, no ejection can be observed with hydrogen plasmas despite similar heat fluxes. Bubble boiling appears to be the main mechanism contributing to the observed droplet ejection.

The construction phase of the linear plasma generator Magnum-PSI at the FOM institute DIFFER has been completed and the facility has been officially opened in March 2012. The scientific program to gain more insight in the plasma–wall interactions relevant for ITER and future fusion reactors has started. In Magnum-PSI, targets of a wide range of materials and shapes can be exposed to high particle, high heat flux plasmas (>1024 ions m−2 s−1; >10 MW/m2). For magnetization of the plasma, oil-cooled electromagnets are temporarily installed to enable pulsed operation until the device is upgraded with a superconducting magnet. The magnets generate a field of up to 1.9 T close to the plasma source for a duration of 6 s. Longer exposure times are available for lower field settings. Plasma characterizations were done with a variety of gases (H, D, He, Ne and Ar) to determine the machine performance and prepare for subsequent scientific experiments. Thomson scattering and optical emission spectroscopy were used to determine the plasma parameters while infrared thermography and target calorimetry were used to determine the power loads to the surface. This paper reports on the status of Magnum-PSI and its diagnostic systems. In addition, an overview of the plasma parameters that can be achieved in the present state will be given.

The combined effect of hydrogen plasma exposure and surface heating, either continuous or by short laser pulses (5 ns), on the surface morphology of rhodium layers has been studied. Investigations were performed by reflectivity measurements, scanning electron microscopy (SEM), X-ray photoelectron spectroscopy (XPS) and atomic force microscope (AFM). While surfaces exposed at room temperature exhibit little modifications, strong surface changes are observed for surface temperatures higher than 250 °C. At 500 °C, the plasma exposed surface exhibits a nanoscale structure (50–100 nm) with a high level of porosity and a low reflectivity. Additional laser irradiation of the surface strongly enhances the observed surface damage. Localized surface melting is observed with craters extending deep into the substrate together with a dense network of voids.

Several laser based diagnostics are implemented on to the linear plasma generator Magnum-PSI, wherein ITER divertor relevant plasma-wall conditions are realized. Laser Induced Desorption Quadrupole Mass Spectroscopy (LID-QMS) and Laser Induced Breakdown Spectroscopy (LIBS) are installed to measure deuterium retention in plasma facing components. Combined with Thermal Desorption Spectroscopy, LID-QMS can be used to measure lateral retention profiles. LIBS is used to measure the surface composition qualitatively, after plasma exposure. An advanced Thomson Scattering (TS) system measures electron density, neutral density and electron temperature profiles (spatial resolution < 2 mm) across the maximum 100 mm plasma diameter. Very low electron density (9 × 10 18 m −3 ) can be measured within seconds with accuracies better than 6%. The minimum measurable electron density and temperature are 1 × 10 17 m −3 and 0.07 eV, respectively. By virtue of the high system sensitivity, single pulse TS can be performed on high density pulsed plasmas (used for replicating ELMs). For measuring the ion temperature and flow velocity of the plasma a Collective TS system (CTS) is being built: the small Debye length of the Magnum-PSI plasma enables application of this method at relatively short laser wavelength. In a feasibility study it was shown that forward CTS with a seeded Nd:YAG laser operating at 1064 nm, can be applied at Magnum-PSI to measure ion temperature and axial velocity with an accuracy of < 8% and < 15%, respectively. Two high spectral resolution (~ 0.005 nm) detection schemes are applied simultaneously: an Echelle grating spectrometer (enabling profile measurements) and a system based on a Fabry-Perot etalon that enables wavelength scanning over its free spectral range, by tilting the device. The status and performance of the various laser based plasma and surface diagnostics will be reported along with experimental results.

Growth of tungsten nano-tendrils (“fuzz”) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The depth of the W fuzz layer (600 ± 150 nm) is consistent with an empirical growth formula from the PISCES experiment. Re-creating the C-Mod exposures as closely as possible in Pilot-PSI experiment can produce nearly-identical nano-tendril morphology and layer thickness at surface temperatures that agree with uncertainties with the C-Mod W probe temperature data. Helium concentrations in W fuzz layers are measured at 1–4 at.%, which is lower than expected for the observed sub-surface voids to be filled with several GPa of helium pressure. This possibly indicates that the void formation is not pressure driven., United States. Dept. of Energy (Award DE-SC00-02060)

Experiments were performed in the Pilot-PSI linear plasma device, to determine the sheath heat transmission coefficients in a high recycling regime under various conditions of density (1–20 × 1020 m−3) and plasma composition (H2, Ar, N2) relevant for the ITER divertor plasma. The 2D surface temperature profile on a tungsten surface was measured with high spatial (0.33 mm) and temporal (200 Hz) resolution using an infrared camera. The target heat flux is calculated using a 2D axis-symmetric Ansys model, the heat transfer is determined from target calorimetry. The plasma parameters are measured with a high resolution Thomson scattering system located 17 mm away from the target surface. Radial profiles of the sheath heat transmission factors can thus be determined. Preliminary results show that γ varies between 4 and 40 depending on the plasma conditions and composition. The value derived from the heat flux calculated with Ansys is significantly lower than theory predicts.

Tritium retention inside the vacuum vessel is a potentially serious constraint in the operation of large-scale fusion machines like ITER. An in situ diagnostics for first wall H/D/T retention by laser induced desorption spectroscopy (LIDS) is proposed for use between plasma discharges. The technique is based on local baking of the first wall by laser irradiation and subsequent analysis of the in-vessel gas by optical emission spectroscopy of plasma radiation. The local heating implementation, kinetics of H/D/T thermal extraction and the accuracy of optical emission spectroscopy measurements are analysed. To resolve the H/D/T lines spectroscopically, their thermal broadening should be minimized to prevent overlapping of the line shapes. A comparative performance analysis of several types of plasma sources with relatively cold ions is made including the following types of discharges: Penning, RF multipactor, laser torch and ECR. All these radiation sources require rather low power and could be used for remote in situ measurements of relative densities of the thermally extracted hydrogen isotopes.

Tungsten button samples were exposed to He ELMing H-mode plasma in DIII-D using 2.3 MW of electron cyclotron heating power. Prior to the exposures, the W buttons were exposed to either He, or D, plasma in PISCES-A for 2000 s at surface temperatures of 225–850 °C to create a variety of surfaces (surface blisters, subsurface nano-bubbles, fuzz). Erosion was spectroscopically measured from each DiMES sample, with the exception of the fuzzy W samples which showed almost undetectable WI emission. Post-exposure grazing incidence small angle x-ray scattering surface analysis showed the formation of 1.5 nm diameter He bubbles in the surface of W buttons after only a single DIII-D (3 s, ~150 ELMs) discharge, similar to the bubble layer resulting from the 2000 s. exposure in PISCES-A. No surface roughening, or damage, was detected on the samples after approximately 600 ELMs with energy density between 0.04–0.1 MJ m−2.

Transient heat pulses with triangular, square, and ELM-like temporal shapes are investigated in order to further understand how transient plasma instabilities will affect plasma facing components in tokamaks. A solution to the 1D heat equation for triangular pulses allows the peak surface temperature to be written analytically for arbitrary rise times. The solution as well as ANSYS simulations reveal that a positive ramp (maximum rise time) triangular pulse has a higher peak surface temperature by a factor of sqrt{2} compared to that from a negative ramp (rise time = 0) pulse shape with equal energy density, peak power, and pulse width. Translating the results to ITER, an ELM or disruption pulse with the shortest rise time is the most benign compared to other pulse shapes with the same peak heat flux and same energy density.

We have performed microstructural characterization using transmission electron microscopy (TEM) techniques to reveal nanometric features in the sub-surface region of tungsten samples exposed to high flux, low energy deuterium plasma. TEM examination revealed formation of a dense dislocation network and dislocation tangles, overall resulting in a strong increase in the dislocation density by at least one order of magnitude as compared to the initial one. Plasma-induced dislocation microstructure vanishes beyond a depth of about 10 mu m from the top of the exposed surface where the dislocation density and its morphology becomes comparable to the reference microstructure. Interstitial edge dislocation loops with Burgers vector a(0)/2 < 111 > and a(0) < 100 > were regularly observed within 6 mu m of the sub-surface region of the exposed samples, but absent in the reference material. The presence of these loops points to a co-existence of nanometric D bubbles, growing by loop punching mechanism, and sub-micron deuterium flakes, resulting in the formation of surface blisters, also observed here by scanning electron microscopy.

Item does not contain fulltext A rare case of aseptic thrombosis of the left vein of Labbé in a young woman is reported. Cerebral venous thrombosis was suggested by computed tomography and confirmed after intravenous administration of contrast by computed tomography venography. The combination of the clinical setting with the findings on the non-enhanced CT may favour the diagnosis of vein of Labbé thrombosis. The diagnosis can be confirmed on computed tomography venography.

Fine-grain graphite targets have been exposed to ITER divertor relevant plasmas in Pilot-PSI to address material migration issues in fusion devices. Optical emission spectroscopy and mass loss measurements have been employed to quantify gross chemical erosion and net erosion yields, respectively. Effects of the ion impact energy and target geometry on carbon erosion yields have been studied. It is concluded that temporal evolution of gross chemical erosion is strongly connected with changes in morphology of plasma exposed surfaces. The net carbon erosion yield is increased when the targets are partly covered by insulating boron-nitride rings., Comment: 14 pages, 4 figures, Contribution to the 19th International Conference on Plasma Surface Interactions

Two cleaning techniques were used for removal of co-deposits from the tested first mirrors exposed in JET: (a) ultrasonic bath; (b) a broad range of polishing conditions from manual buffing to machine polishing with the diamond grain size of up to 3 μm. Reflectivity measurements were performed after each step in the cleaning procedure. Surfaces were also examined with electron microscopy and ion beam analysis methods. Ultrasonic cleaning leads to partial recovery of reflectivity due to enhanced detachment of deposits. Typically 30–50% of the original reflectivity was recovered in the visible light and 50–90% in the infrared region. One mirror was cleaned completely. Polishing with diamond paste may lead to successful removal of deposits but the damage to the surface in case of the large diamond grains was observed. Recovery of up to 100% of the initial reflectivity was achieved for some mirrors.

The formation of metallic nanostructures by exposure of molybdenum and tungsten surfaces to high fluxes of low energy helium ions is studied as a function of the ion energy, plasma exposure time, and surface temperature. Helium plasma exposure leads to the formation of nanoscopic filaments on the surface of both metals. The size of the helium-induced nanostructure increases with increasing surface temperature while the thickness of the modified layer increases with time. In addition, the growth rate of the nanostructured layer also depends on the surface temperature. The size of the nanostructure appears linked with the size of the near-surface voids induced by the low energy ions. The results presented here thus demonstrate that surface processing by low-energy helium ions provides an efficient route for the formation of porous metallic nanostructures. (C) 2012 American Vacuum Society.

For understanding carbon erosion and redeposition in nuclear fusion devices, it is important to understand the transport and chemical break-up of hydrocarbon molecules in edge plasmas, often diagnosed by emission of the CH A(2)Delta-X-2 Pi Gero band around 430 nm. The CH A-level can be excited either by electron-impact (EI) or by dissociative recombination (DR) of hydrocarbon ions. These processes were included in the 3D Monte Carlo impurity transport code ERO. A series of methane injection experiments was performed in the high-density, low-temperature linear plasma generator Pilot-PSI, and simulated emission intensity profiles were benchmarked against these experiments. It was confirmed that excitation by DR dominates at T-e < 1.5 eV. The results indicate that the fraction of DR events that lead to a CH radical in the A-level and consequent photon emission is at least 10%. Additionally, quenching of the excited CH radicals by EI de-excitation was included in the modeling. This quenching is shown to be significant: depending on the electron density, it reduces the effective CH emission by a factor of 1.4 at n(e) = 1.3 x 10(20) m(-3), to 2.8 at n(e) = 9.3 x 10(20) m(-3). Its inclusion significantly improved agreement between experiment and modeling.

A set of stainless steel (SS) and molybdenum mirror samples located in the divertor and at the outer mid-plane of the vessel were exposed in JET from 2005 to 2007. A selection of these mirror samples with well adhered deposits (i.e. not flaking) of up to a few hundred nanometers in thickness and with Be/C ratios ranging from 0 to similar to 1 have been cleaned using a laser system developed at CEA, Saclay. Following laser cleaning the recovered reflectivity was generally better in the infrared than the visible spectrum, with recovery of up to 90% of the initial reflectivity being obtained at 1600 nm for both Mo and SS mirrors falling as low as 20-30% of initial reflectivity at a wavelength of 400 nm for some SS mirrors, rising to similar to 80% for Mo mirrors. Some deposit remained on the mirrors after the cleaning trials. (C) 2010 EURATOM. Published by Elsevier B.V. All rights reserved.

Among dust particles of various origins collected in the MAST vessel, large quantities of carbon nanoparticles are observed. They are produced in the coldest plasma regions from the nucleation of carbonaceous species originating from physical and chemical erosion of the graphitic plasma facing components. On the other hand, laboratory plasmas studies show that in an environment of reactive carbonaceous species, complex chemical pathways lead to various molecular precursors which depend strongly on plasma conditions. The latter vary very often in a tokamak. Nevertheless, the final nanoparticles produced in laboratory plasmas and in MAST present similarities, giving indications on common growth processes in the plasma.

The first mirror test for ITER in JET with carbon walls has been completed. Thirty polycrystalline Mo mirrors including four coated with a 1 μm rhodium (Rh) film were exposed to plasma in the divertor region and in the main chamber. The mirrors were installed in eight cassettes of pan-pipe shape. The reflectivity of all mirrors exposed in the divertor has been degraded by 80–90% because of the formation of thick (>20 μm) flaking co-deposits on surfaces. Only small reflectivity losses (5–10%) occurred on mirrors located at the channel mouth of the cassettes from the main chamber wall. This is due to the in situ removal of deposited species by charge exchange neutrals. Deuterium, 12C and 9Be are the main isotopes detected on surfaces, but other isotopes (13C) are also found in some locations, thus indicating differences in the material migration. Rhodium coatings with an initial reflectivity that is 30% better than that of pure Mo survived the test without detachment, but their resultant reflectivity was the same as that of the exposed Mo surfaces.

Thick free-standing boron-doped diamonds were prepared by microwave plasma assisted chemical vapour deposition. Samples with a final thickness close to 5 mm and with lateral dimensions 25 × 25 mm were produced. The thermal shock resistance of the material was tested by exposure in the JUDITH electron gun to reproduce the conditions expected in a fusion reactor during plasma instabilities. Boron-doped diamond is shown to exhibit thermal shock resistance far better than current material of choice for a fusion reactor like tungsten, with no surface damage after exposure to 100 cycles at 2.5 GW m−2 for 5 ms.

In future fusion devices like ITER deposition of impurities will likely occur in areas, remote from plasma and on the sensitive components of optical diagnostics, like mirrors and windows. Deposition in remote areas may lead to the tritium retention and therefore represent a safety issue. Deposition on optical components will severely deteriorate their optical properties potentially leading to the shutdown of the respective diagnostic systems. An active control over deposition is therefore highly desirable. The paper contains an overview of experiments on active control over carbon deposition undertaken in TEXTOR tokamak. A prototype of diagnostic duct was exposed in the scrape-off layer plasmas of TEXTOR where several techniques were applied to mitigate carbon deposition on diagnostic mirrors located inside this prototype. The complete suppression of carbon deposition on the surface of diagnostic mirror was achieved by feeding the deuterium gas inside the diagnostic duct. (C) 2010 Elsevier B.V. All rights reserved.

Several types of diamond layers have been deposited on molybdenum tiles by chemical vapour deposition techniques, and exposed under erosion-dominated conditions in the SOL of TEXTOR in order to assess them as a suitable candidate for plasma-facing material. Post-exposure characterisation of physical properties and surface modification induced by the plasma was performed by SEM imaging, investigation of diamond surface by micro-Raman spectroscopy and deuterium retention measurements by NRA. The analyses evidenced that lightly boron-doped micro-crystalline diamond is performing better than undoped and heavily doped samples, and nano-crystalline diamond and diamond-like carbon, as it showed lower surface modification, lower presence of arcing traces at the surface and lower deuterium retention. High concentration of boron in the layers led to higher retention of deuterium, whereas undoped (insulating) diamond showed increased arcing activity. Nano-crystalline diamond and diamond-like carbon layers generally showed poorer mechanical properties. (C) 2010 Elsevier B.V. All rights reserved.

We report on the generation of high transient heat and particle fluxes in a linear plasma device by pulsed operation of the plasma source. A capacitor bank is discharged into the source to transiently increase the discharge current up to 1.7 kA, allowing peak densities and temperature of 70x10(20) m(-3) and 6 eV corresponding to a surface power density of about 400 MW m(-2). (C) 2010 American Institute of Physics. [doi: 10.1063/1.3484961]

Microcrystalline undoped and heavily boron-doped polycrystalline diamond layers have been deposited on various substrates by hot filament CVD and exposed to hydrogen plasma in a linear plasma reactor (Pilot-PSI, The Netherlands) that simulates the high flux and high density plasma conditions of tokamak divertors, as well as in the DIII-D tokamak (US). Pre- and post-exposure analysis by SEM and Raman spectroscopy characterised the surface appearance and the sp3 and sp2 components of the diamond films respectively. Surface roughness variation was measured by AFM. Hardness and Young's modulus were assessed by nanoindentation in order to characterise the effect of the plasma on the mechanical properties. HRTEM and EELS have been used to evaluate the nature of the modification induced at the diamond surface by the plasma exposure. The measurements have shown that, despite some surface amorphisation of the exposed layers, further long-term exposure studies are warranted as neither delamination, dramatic film failure nor entire erosion of the film were observed. © 2010 Elsevier B.V. All rights reserved.

Dust samples collected in the MAST tokamak have been characterized. Mass measurements have been correlated to regions of collection. Electron microscopy has revealed the presence of large quantities of carbon nanoparticles produced in gas phase as well as the presence of rolled-up carbon thin layers whatever the collection region. Shape, structure and chemical composition have been established by means of complementary diagnostics such as electron microscopy, electron diffraction, energy dispersed X-ray spectroscopy, micro-Raman spectroscopy and infrared absorption spectroscopy. (C) 2010 Elsevier B.V. All rights reserved.

A long wavelength range infrared (LWIR) camera has been installed on MAST to compliment the existing medium wave infrared (MWIR) system. Simultaneous LWIR/MWIR temperature measurements have been made in the lower and upper divertors of MAST. As expected, the LWIR system is less sensitive to disturbances from hot spots on the surface. A thorough correction of the surface effects using both systems has improved the reliability of the heat flux calculations in MAST. A careful assessment of the heat load asymmetries as a function of the magnetic geometry has been carried out during a variety of L-mode discharges (single and double null, reversed and forward field). In all cases, most of the energy arriving in the scrape-off layer goes to the outer divertor. The in/out power ratio is found to be different for lower and upper single-null discharges with more power going to the outer divertor in the latter case. In addition, it is found that the heat flux profile width is different in the upper and lower divertor. The heat flux profile width at the secondary strike point increases with increasing vertical bar delta r(sep)vertical bar.

We have studied the chemical erosion of different carbon composites in Pilot-PSI at ITER-relevant hydrogen plasma fluxes (∼1024 m -2 s-1) and low electron temperatures (Te∼1 eV). Optical emission spectroscopy on the CH A-X band was used to characterize the chemical sputtering. Fine grain graphite (R 6650, SGL Carbon Group), ITER-reference carbon fiber composite material (SNECMA NB31 and NB41; Dunlop 3D), nano- and micro-crystalline diamond coatings on molybdenum and SiC (Silit® SKD Reaction-Bonded, Saint-Gobain Ceramics) were compared. The chemical sputtering was similar for the different composites under comparable plasma conditions, except for SiC, which produced a ten times lower rate. The CH emission was constant at electron temperatures Te>1 eV and ion fluxes ranging between 1023 and 1024 m- 2 s-1, but decreased at lower temperatures. This decrease is possibly due to changes in the excitation of CH and not due to a change in the chemical erosion rate.

The levels of retention in codeposited layers of each of the three ITER materials (C, Be and W) are compared. Scaling laws, based on the conditions during the codeposition process (surface temperature, incident particle energy and ratio of the depositing fluxes), are presented to allow prediction of expected retention under ITER conditions. Retention in carbon codeposits scales inversely with incident particle energy, whereas in the metallic codeposits the retention level scales proportionally to increasing particle energy. The differing scaling of retention with incident particle energy provides insights into which material may impact the global retention in ITER depending on where it may form codeposits. In addition to the amount of retention, the release behaviour of tritium from codeposits will influence the tritium accumulation rate within ITER. The thermal release behaviour of T (or D) from codeposits can be used to evaluate the effectiveness of baking at different temperatures as a means of tritium removal. Finally, the desorption kinetics from Be and W codeposits are contrasted. In the case of W codeposits, the duration of the baking cycle is important in determining the removal efficiency, whereas with Be codeposited layers, the maximum achievable bake temperature plays the leading role in determining removal efficiency.

Different mechanisms contribute to tritium retention in ITER, amongst which co-deposition with materials from the plasma-facing components is one of the main contributors. A systematic study of the influence of the deposition conditions (substrate temperature, deposition rate, energy of the incident particles) on the deuterium retention in co-deposited beryllium layers has been carried out in PISCES-B. The mechanism by which deuterium co-deposits with beryllium appears to be a combination of co-deposition and implantation, with a decreased retention for increased deposition rate and an increased retention for increased incident deuterium particle energy. A scaling equation is developed, providing a method to predict the retention in Be co-deposits formed in PISCES-B as a function of the layer formation conditions. Using this equation, previously published data on retention in Be co-deposits are re-examined and relatively good agreement is found with the prediction of the scaling equation.

In this paper, the results of postmortem analysis of mirror samples exposed for the one-year campaign inside the Tore Supra (TS) tokamak and for shorter time inside the TCV tokamak are presented, together with results of laboratory mirror experiments. In TS, although all mirrors were eroded, the observed erosion rates do not correspond with the published sputtering yields. In TCV the thickness of the carbon-based film deposited on the mirror samples was found to depend on the mirror material, being several times higher on Si than on W and Mo substrates. An attempt is made to attract for explanation of results the difference in stickiness of carbon to SS and to Cu surfaces. The laboratory experiments qualitatively support such explanation.

Metallic mirrors will be essential components of the optical diagnostic systems in the International Thermonuclear Experimental Reactor (ITER). Reliability of these systems may be affected by mirror reflectivity changes induced by erosion and/or deposition of impurities (carbon, beryllium). The present study aims to assess the effect of beryllium (Be) deposition on the reflectivity of metallic mirrors and to collect data on the optical quality of these layers in terms of morphology, roughness, etc. Mirrors from molybdenum and copper were exposed in the PISCES-B linear plasma device to collect eroded material from graphite and beryllium targets exposed to beryllium-seeded deuterium plasma. After exposure, relative reflectivity of the mirrors was measured and different surface analysis techniques were used to investigate the properties of the deposited layers. Be layers formed in PISCES-B exhibit high levels of porosity which makes the reflectivity of the Be layers much lower than the reflectivity of pure Be. It is found that if Be deposition occurs on ITER first mirrors, the reflectivity of the coated mirrors will strongly depend on the layer morphology, which in turn depends on the deposition conditions. (C) 2007 American Institute of Physics.

Metallic mirrors will be essential components of all optical spectroscopy and imaging systems for ITER plasma diagnostics. Any change in the mirror performance, in particular, its reflectivity, due to erosion of the surface by charge exchange neutrals or deposition of impurities will influence the quality and reliability of the detected signals. Due to its high reflectivity in the visible wavelength range and its low sputtering yield, rhodium appears as an attractive material for first mirrors in ITER. However, the very high price of the raw material calls for using it in the form of a film deposited onto metallic substrates. The development of a reliable technique for the preparation of high reflectivity rhodium films is therefore of the highest importance. Rhodium layers with thicknesses of up to 2 mu m were produced on different substrates of interest (Mo, stainless steel, Cu) by magnetron sputtering. Produced films exhibit a low roughness and crystallite size of about 10 nm with a dense columnar structure. No impurities were detected on the surface after deposition. Scratch tests demonstrate that adhesion properties increase with substrate hardness. Detailed optical characterizations of Rh-coated mirrors as well as results of erosion tests performed both under laboratory conditions and in the TEXTOR tokamak are presented in this paper. c 2007 American Institute of Physics.

The plasma-facing components (PFCs) in future fusion devices such as ITER will receive intense transient heat fluxes from plasma instabilities, such as edge localized modes (ELMs) and disruptions, which will limit material lifetime. The energy density and transient pulse duration are typically used to characterize the PFC damage threshold. However, these parameters are not sufficient to define a damage threshold, because different transient pulse shapes with the same energy density and same pulse duration produce different peak surface temperatures (and thus stresses, which ultimately determine material damage). The surface temperature and damage of tungsten (the material to be used for the ITER divertor target armor) in the form of surface roughening and melting are investigated using various temporal pulse shapes from an Nd:YAG laser in the PISCES-B facility. The heat flux factor is examined and shown to be an inadequate parameter to characterize the temperature rise except for square temporal pulse shapes. For ELM-like heat pulses, the long tail in the temporal shape results in a lower peak surface temperature and less damage compared to a symmetric triangle pulse with equal energy density.

Metallic mirrors are foreseen in ITER diagnostic systems as optical elements directly viewing the plasma radiation. In the frame of an EFDA contract, metallic mirror samples have been exposed for long pulse plasma discharges in Tore Supra (TS) in order to investigate surface modifications caused by erosion and re-deposition processes. Three different materials have been selected: mono-crystalline molybdenum (mc-Mo), polycrystalline stainless steel (SS) and copper (Cu). The mc-Mo samples showed after TS exposure almost no surface roughness modifications and the lowest net-erosion. A slight reflectivity reduction, most pronounced in the near UV, is attributed to light absorption in a thin carbon deposit. Cu mirrors showed by far the highest surface roughness, erosion and diffusive reflectivity. Comparative laboratory glow discharge experiments with virgin reference samples and numerical simulations of erosion/deposition confirm the dominant contribution of conditioning procedures to erosion of mirrors exposed (without shutter protection) in Tore Supra.

Metallic mirrors will be used in ITER for optical diagnostics working in different spectral ranges. Their optical properties will change with time due to erosion, deposition, and particle implantation. First tests of molybdenum mirrors were performed in the DIII-D divertor under deposition-dominated conditions. Two sets of mirrors recessed 2 cm below the divertor floor in the private flux region were exposed to a series of identical, lower-single-null, ELMing (featuring edge localized modes) H-mode discharges with detached plasma conditions in both divertor legs. The first set of mirrors was exposed at ambient temperature, while the second set was preheated to temperatures between 140 and 80 S C. During the exposures mirrors in both sets were additionally heated by radiation from the plasma. The nonheated mirrors exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. Net carbon deposition rate on the preheated mirrors was a factor of 30 - 100 lower and their optical reflectivity in the wave range above 500 nm was essentially preserved. (c) 2006 American Institute of Physics.

Rotary fixation of the atlanto-axial joint is a rare disorder in which the atlas gets fixed in a position normally achieved during rotation. A persistent asymmetrical position of the dens in relation to the atlas during head rotation is the clue to the diagnosis. Rotary fixation can be demonstrated by radiography/fluoroscopy and computed tomography (CT). However it is easily overlooked, largely by inattention. We present a typical case in an adult in whom the rotary fixation was demonstrated by radiography/fluoroscopy and CT.

Pelvic inflammatory disease with tuboovarian abscess is rare in postmenopausal women. Clinical and sonographic findings are usually sufficient to recognize pelvic inflammatory disease in premenopausal women, but in the elderly the disease may easily be overlooked, largely by unexpectedness. Computed tomography can be helpful when the clinical and sonographic findings are complex or equivocal. However, when the level of suspicion is low, it can be very difficult to interpret, especially when complicated with peritonitis, as illustrated in the present case report.

The snowflake (SF) divertor is a plasma configuration that may enable tokamak operation at high performance and lower peak heat loads on the plasma-facing components than a standard single-null divertor. This paper reports on the results of experiments performed on the TCV tokamak in both the low- and high-confinement regimes, wherein the divertor configuration was continuously varied between a standard single-null and a 'SF-plus', which features auxiliary strike points (SPs) in the private flux region of the primary separatrix. The measured edge properties show that, in L-mode, the fraction of the exhaust power reaching the additional SPs is small. During edge-localized modes, up to similar to 20% of the exhausted energy is redistributed to the additional SPs even at an x-point separation of 0.6 times the plasma minor radius, thereby reducing the peak heat flux to the inner primary SP by a factor of 2-3. The observed behaviour is qualitatively consistent with a proposed model for enhanced cross-field transport through the SF's relatively large region of low poloidal field by instability-driven convection.

Transport of hydrocarbon impurities in a high-density (>10 20 m−3), low-temperature (