[Show abstract][Hide abstract] ABSTRACT: Thermo-chemical removal (TCR), or baking in reactive gases, is a candidate method to control the co-deposit related tritium inventory in fusion devices. TCR can be understood as reaction–diffusion processes in a porous material. O2-TCR was applied to 150–550 nm thick a-C:D layers with similar textures. A linear relation between the integral TCR rate and the layer thickness, as predicted by the understanding, was observed in the experiment, i.e. the time to remove the hydrogen inventory is independent of its initial amount. TCR with nitrogen dioxide (NO2) at temperatures of 200–350 °C was conducted with a set of a-C:D and W–C–H layers. At 350 °C NO2 removed ~ 15% porosity a-C:D within 3 min. The O retention in remaining a-C:D was ≈ 1017 O cm−2. An activation energy of ≈ 0.78 eV for reactions of NO2 with D and C was determined. The results were applied for predictions of the TCR effectivity in ITER. The treatment of W–C–H led to O uptake (O/W ≈ 2–3), while W and C contents remained unchanged.
Physica Scripta 04/2014; 2014(T159):014065. · 1.03 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s−1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas.
Physica Scripta 04/2014; 2014(T159):014063. · 1.03 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The ITER-like wall recently installed in JET comprises solid beryllium limiters and a combination of bulk tungsten and tungsten-coated carbon fibre composite divertor tiles without active cooling. During a beryllium power handling qualification experiment performed in limiter configuration with 5 MW neutral beam injection input power, accidental beryllium melt events, melt layer motion and splashing were observed locally on a few beryllium limiters in the plasma contact areas. The Lorentz force is responsible for the observed melt layer movement. To move liquid beryllium against the gravity force, the current flowing from the plasma perpendicularly to the limiter surface must be higher than 6 kA m−2. The thermo-emission current at the melting point of beryllium is much lower. The upward motion of the liquid beryllium against gravity can be due to a combination of the Lorentz force from the secondary electron emission and plasma pressure force.
Physica Scripta 04/2014; 2014(T159):014041. · 1.03 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Cracking thresholds and crack patterns in tungsten targets after repetitive ITER-like edge localized mode (ELM) pulses have been studied in recent simulation experiments by laser irradiation. The tungsten specimens were tested under selected conditions to quantify the thermal shock response. A Nd:YAG laser capable of delivering up to 32 J of energy per pulse with a duration of 1 ms at the fundamental wavelength λ = 1064 nm has been used to irradiate ITER-grade tungsten samples with repetitive heat loads. The laser exposures were performed for targets at room temperature (RT) as well as for targets preheated to 400 °C to measure the effects of the ELM-like loading conditions on the formation and development of cracks. The magnitude of the heat loads was 0.19, 0.38, 0.76 and 0.90 MJ m−2 (below the melting threshold) with a pulse duration of 1 ms. The tungsten surface was analysed after 100 and 1000 laser pulses to investigate the influence of material modification by plasma exposures on the cracking threshold. The observed damage threshold for ITER-grade W lies between 0.38 and 0.76 GW m−2. Continued cycling up to 1000 pulses at RT results in enhanced erosion of crack edges and crack edge melting. At the base temperature of 400 °C, the formation of cracks is suppressed.
Physica Scripta 04/2014; 2014(T159):014005. · 1.03 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The control of the radioactive inventory in the vacuum vessel of ITER is a main safety issue. Erosion of activated plasma-facing components (PFC) and co-deposition of tritiated dust on PFC and in areas below the divertor constitute the main sources of in-vessel radioactive inventory mobilizable in the case of an accident and also during venting of the vessel. To trace the dust and tritium inventory in the machine, the use of collectors in the form of removable samples was evaluated, beside other techniques, since it provides a reliable way to follow the history of the deposits and check critical areas. Four types of removable probes and two optional active diagnostics were selected out of about 30 different options. For all four probes, a conceptual design was worked out and the feasibility was checked with preliminary estimations of thermal and electromagnetic loads, as well as remote handling paths. The highest temperature estimated for the front face of all probes lies in the range 300–500 °C, which is tolerable. Installed in representative places, such removable samples may provide information about the dust and tritium distribution inside the vacuum vessel.
Physica Scripta 04/2014; 2014(T159):014004. · 1.03 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: First results from three-dimensional modeling of the divertor heat and particle flux pattern during application of resonant magnetic perturbation fields as ELM control scheme in ITER with the EMC3-Eirene fluid plasma and kinetic neutral transport code are discussed. The formation of a helical magnetic footprint breaks the toroidal symmetry of the heat and particle fluxes. Expansion of the flux pattern as far as 60 cm away from the unperturbed strike line is seen with vacuum RMP fields, resulting in a preferable heat flux spreading. Inclusion of plasma response reduces the radial extension of the heat and particle fluxes and results in a heat flux peaking closer to the unperturbed level. A strong reduction of the particle confinement is found. 3D flow channels are identified as a consistent reason due to direct parallel outflow from inside of the separatrix. Their radial inward expansion and hence the level of particle pump out is shown to be dependent on the perturbation level.
Journal of Nuclear Materials 07/2013; 438:S194–S198. · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Molecular spectroscopy was used to observe molecular deuterium at the
outer strike point of the new bulk tungsten JET divertor. The rotational
and vibrational populations of the deuterium molecules in the ground
state were determined from the deuterium Q-branches of Fulcher-α
band emission (d3Πu-→a3Σg+) in the
600-640 nm spectral range. For L-mode plasmas in the low recycling
regime the molecular emission maximum is located in the vicinity of the
strike point. The spatial profile of the emission was strongly modified
during plasma detachment in both L- and H-mode plasmas. The rotational
temperature of excited molecules reached 2760 K in L-mode. The
vibrational population has a peculiarity: a remarkably high population
of the d3Πu-(v = 0) vibrational level indicating a
non-Boltzmann vibrational distribution of D2 in tungsten
Journal of Nuclear Materials 07/2013; · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Resonant Magnetic Perturbations (RMPs) are applied with the Dynamic
Ergodic Divertor (DED) at TEXTOR to control the plasma edge transport
and the plasma surface interaction. This leads to the formation of a
three-dimensional (3D) topology of the scrape-off layer (SOL). To
quantify the erosion/deposition balance and the material migration in
this 3D boundary, spherical test limiters were exposed to plasmas with
and without RMP fields applied. Methane doped with 13C as
tracer element was injected through a gas inlet in the test limiter. The
local gas source was monitored by spatially resolving spectroscopy and
the resulting deposition patterns on the limiters were analysed with
colourimetry and nuclear reaction analysis. These measurements were
compared to simulations of the magnetic field topology simulations. The
data provide evidence of a particle migration dominated by an ExB drift
within stochastic zones of the 3D plasma boundary.
Journal of Nuclear Materials 07/2013; · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: L-mode and H-mode density limits with the ITER-like wall (ILW) have been
investigated in the recent experimental campaign and compared with
experiments in the JET carbon material configuration. The density limit
is up to 40% higher in the JET-ILW than in the JET-CFC machine. This is
linked to the formerly higher radiation fraction and, correspondingly,
to earlier divertor detachment in the JET-CFC. In the ILW configuration,
the discharge demonstrates a stable operation with a completely detached
outer divertor in L- and H-mode. In contrary to the well-known "heating
power independent" Greenwald limit, the L-mode densities limit increases
moderately with rising heating power (˜Pheat0.4) independently of
the wall material.The H-L transition constitutes an effective
undisruptive density limit for an H-mode plasma. Detachment itself does
not trigger the H-L back transition and does not present a limit on
plasma density. In the range of neutral beam heating 8-10.5 MW, no
dependence of the H-mode density limit on the heating power was
Journal of Nuclear Materials 07/2013; 438:S139–S147. · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: On the TEXTOR tokamak various experiments aimed at investigation of tungsten erosion and transport are performed. In one experiment a spherical W/C twin limiter positioned close to the last-closed flux surface in the near scrape-off layer was exposed to a number of comparable plasma discharges with stepwise variations of edge plasma parameters. Spatial distribution of tungsten and carbon light emission was recorded with two dimensional CCD cameras and spectrometer systems with high spectral and spatial resolution. Penetration depths, tungsten sputtering fluxes and erosion yields were measured. Comparison between experimental data and the results of modelling with the 3D Monte-Carlo code ERO is performed. The main objective of this study was to test the adequacy of the existing atomic data for neutral tungsten. The modelled penetration depths of the light emission of tungsten are a factor of 2–3 smaller than in experiment, which may indicate the overestimation of ionization rates.
Journal of Nuclear Materials 07/2013; 438:S351–S355. · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: In the linear plasma simulator PSI-2, capable of producing electron
temperatures and densities of 1-20 eV and
1017-1019 m‑3 in the target
region, a deuterium plasma with a simultaneous ionising and recombining
region has been produced. The latter appeared when a sufficiently high
extra deuterium gas flow was applied at the target region. An imaging
spectrometer was used to measure Te and ne -
profiles via the analysis of the Paschen-series resulting in drops down
to 0.1 eV and 2 × 1018 m‑3 in the
recombining region whereas intensity ratios of Hα, Hβ and
Hγ provided ne in the ionising part. In the latter case
ne was also deduced from the rotational temperature of the
D2 molecules. Comparison with probe data yielded a reasonable
agreement in that case. A detailed analysis of the atomic level
populations and their light emissions provided some insight into the
contribution of possible recombination mechanisms such as EIR and MAR.
Journal of Nuclear Materials 07/2013; · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: In ITER, mirrors will be used as plasma-viewing elements in all optical and laser diagnostics. In the harsh environment, mirror performance will degrade hampering the operation of associated diagnostics. The most adverse effect on mirror performance is caused by the deposition of impurities. It is expected that the most challenging situation will occur in the divertor. With the envisaged changes to all-metal plasma-facing components (PFCs) in ITER, an assessment of mirror performance in an existing divertor tokamak with all-metal PFCs is urgently needed. Molybdenum and copper mirrors were exposed for nearly nine months in ASDEX Upgrade which has all-tungsten PFCs. Mirrors were located at the inner wall, under the dome and in the pump duct. During exposure, the mirrors were heated to temperature in the range 145–165 • C. This was made to approach the expected level of heating due to absorption of neutrons and gammas on mirrors in the ITER divertor. After exposure, degradation of the reflectivity was detected on all mirrors. The highest reflectivity drop was measured on mirrors under the dome facing the outer strike point, reaching −55% at 500 nm. The least degradation was detected on mirrors in the pump duct, where the reflectivity was preserved in the range 500–2500 nm and the largest decrease was about −8% at 250 nm. On all contaminated mirrors carbon fraction did not exceed 50 at% while the major contaminants were metals and oxygen. The degradation of exposed mirrors underlines the necessity for urgent R&D on deposition mitigation and in situ mirror cleaning in ITER. (Some figures may appear in colour only in the online journal)
[Show abstract][Hide abstract] ABSTRACT: Laser-based methods are investigated for the development of an in situ diagnostic for spatially and temporally resolved characterization of the first wall in fusion devices. Here we report on the first systematic laser-induced ablation spectroscopy (LIAS) measurements carried out on various surface layers in the TEXTOR tokamak. These materials include a-C:D, mixed W/C/Al/D2, Oerlikon Balzers ‘Balinit’ diamond-like carbon layers and EK98 fine-grain graphite. In LIAS, the bulk or deposited material is evaporated during the plasma discharge by intense laser radiation. The light emitted by particles entering the edge of the ionizing tokamak plasma is then observed by optical spectroscopy. In the measurements taken, it was found that the studied layers can be identified by their characteristic line emission. A good correlation between the observed line intensity and layer thickness is found. The observed plumes show target material dependence. To analyze layers formed during tokamak operation, further investigation of the ablation process and reference materials for cross calibration is required.
[Show abstract][Hide abstract] ABSTRACT: This contribution gives an overview of different simulation methods for transient events and damages induced in tungsten. The investigations were focussed on the resulting crack networks and special attention was paid to crack distance, width and depth. The results indicate that the different techniques show, in general, similar damage behaviours and the same damage thresholds.
Fusion Science and Technology 01/2013; 63(1T):197-200. · 0.52 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Development of methods to characterise the first wall in ITER and future fusion devices without removal of wall tiles is important to support safety assessments for tritium retention and dust production and to understand plasma wall processes in general. Laser based techniques are presently under investigation to provide these requirements, among which Laser Induced Desorption Spectroscopy (LIDS) is proposed to measure the deuterium and tritium load of the plasma facing surfaces by thermal desorption and spectroscopic detection of the desorbed fuel in the edge of the fusion plasma. The method relies on its capability to desorb the hydrogen isotopes in a laser heated spot. The application of LID on bulk tungsten targets exposed to a wide range of deuterium fluxes, fluences and impact energies under different surface temperatures is investigated in this paper. The results are compared with Thermal Desorption Spectrometry (TDS), Nuclear Reaction Analysis (NRA) and a diffusion model.
[Show abstract][Hide abstract] ABSTRACT: JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments ( I p = 2.0 MA, B t = 2.0–2.4 T, δ = 0.2–0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system is reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges ( P aux = 0.5 MW), type III ( P aux = 5.0 MW) and type-I ELMy H-mode plasmas ( P aux = 12.0 MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27 × 10 21 , 1.37 × 10 21 and 1.97 × 10 21 D s −1 down to 4.8 × 10 19 , 7.2 × 10 19 and 16 × 10 19 D s −1 , respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5 × 10 20 D s −1 is obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device.
[Show abstract][Hide abstract] ABSTRACT: Tracer experiments have been carried out by injection 13C marked methane through test limiters exposed to the scrape-off-layer in TEXTOR. The influence of impact energy and flux on depositing 13C species has been studied. One experiment has been performed with biased test limiter (−300 V) in order to increase energy of positively charged ions and the other one with 10 times reduced 13CH4 injection rate compared to previously used injection rate. Biasing of the test limiter increases the resulting 13C deposition by a factor of ∼6 – post-mortem analysis yields a 13C deposition efficiency of ∼1.7% compared to ∼0.3% without biasing. Reducing the injection rate increases 13C deposition efficiency to ∼0.7%, which is more than two times larger compared to experiments with previously used injection rate. ERO modelling shows that enhanced re-erosion of redeposits is still necessary to reproduce measured 13C deposition efficiencies.
Journal of Nuclear Materials 01/2013; · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: A new endoscope with optimised divertor view has been developed in order to survey and monitor the emission of specific impurities such as tungsten and the remaining carbon as well as beryllium in the tungsten divertor of JET after the implementation of the ITER-like wall in 2011. The endoscope is a prototype for testing an ITER relevant design concept based on reflective optics only. It may be subject to high neutron fluxes as expected in ITER. The operating wavelength range, from 390 nm to 2500 nm, allows the measurements of the emission of all expected impurities (W I, Be II, C I, C II, C III) with high optical transmittance (≥30% in the designed wavelength range) as well as high spatial resolution that is ≤2 mm at the object plane and ≤3 mm for the full depth of field (±0.7 m). The new optical design includes options for in situ calibration of the endoscope transmittance during the experimental campaign, which allows the continuous tracing of possible transmittance degradation with time due to impurity deposition and erosion by fast neutral particles. In parallel to the new optical design, a new type of possibly ITER relevant shutter system based on pneumatic techniques has been developed and integrated into the endoscope head. The endoscope is equipped with four digital CCD cameras, each combined with two filter wheels for narrow band interference and neutral density filters. Additionally, two protection cameras in the λ > 0.95 μm range have been integrated in the optical design for the real time wall protection during the plasma operation of JET.
The Review of scientific instruments 10/2012; 83(10):10D511. · 1.52 Impact Factor