[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: In order to evaluate the applicability of toughened, fine-grained, recrystallized (TFGR)-W to tokamak edge plasma environment, two TFGR-W specimens (TFGR-W 1.1wt%TiC and TFGR-W 3.3wt%TaC) were exposed to 31 identical ohmic discharges in the TEXTOR tokamak by means of a limiter lock system. The highest surface temperature reached was about 1300 °C. Under these temperature conditions, the bulk microstructure and dispersoids distribution of both TFGR-W remained intact, suggesting that these TFGR-W tungsten materials have sufficient stability under these plasma loading conditions. The erosion of TiC dispersoids on the surface was enhanced by plasma exposure above 1150 °C, while such enhanced erosion was not observed for TaC dispersoids probably due to the higher melting temperature of Ta than Ti.
Physica Scripta 04/2014; 2014(T159):014038. · 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: 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: This review summarizes surface morphology changes of tungsten caused by heat and particle loadings from edge plasmas, and their effects on enhanced erosion and material lifetime in ITER and beyond. Pulsed heat loadings by transients (disruption and ELM) are the largest concerns due to surface melting, cracking, and dust formation. Hydrogen induced blistering is unlikely to be an issue of ITER. Helium bombardment would cause surface morphology changes such as W fuzz, He holes, and nanometric bubble layers, which could lead to enhanced erosion (e.g. unipolar arcing of W fuzz). Particle loadings could enhance pulsed heat effects (cracking and erosion) due to surface layer embrittlement by nanometric bubbles and solute atoms. But pulsed heat loadings alleviate surfaces morphology changes in some cases (He holes by ELM-like heat pulses). Effects of extremely high fluence (∼1030 m−2), mixed materials, and neutron irradiation are important issues to be pursued for ITER and beyond. In addition, surface refurbishment to prolong material lifetime is also an important issue.
[Show abstract][Hide abstract] ABSTRACT: The analysis of the deposition of eroded wall material on the plasma-facing materials in fusion devices is one of the crucial issues to maintain the plasma performance and to fulfill safety requirements with respect to tritium retention by co-deposition. Laser ablation with minimal damage to the plasma facing material is a promising method for in situ monitoring and removal of the deposition, especially for plasma-shadowed areas which are difficult to reach by other cleaning methods like plasma discharge. It requires the information of ablation process and the ablation threshold for quantitative analysis and effective removal of the different deposits. This paper presents systemic laboratory experimental analysis of the behavior of the ITER relevant materials, graphite, tungsten, aluminum (as a substitution of beryllium) and mixed deposits ablated by a Nd:YAG laser (1064 nm) with different energy densities (1–27 J/cm2, power density 0.3–3.9 GW/cm2). The mixed deposits consisted of W–Al–C layer were deposited on W substrate by magnetron sputtering and arc plasma deposition. The aim was to select the proper parameters for the quantitative analysis and for laser removal of the deposits by investigating the ablation efficiency and ablation threshold for the bulk materials and deposits. The comparison of the ablation and saturation energy thresholds for pure and mixed materials shows that the ablation threshold of the mixed layer depends on the concentration of the components. We propose laser induced breakdown spectroscopy for determination of the elemental composition of deposits and then we select the laser parameters for the layer removal. Comparison of quantitative analysis results from laboratory to that from TEXTOR shows reasonable agreements. The dependence of the spectra on plasma parameters and ambient gas pressure is investigated.
Journal of Nuclear Materials. 01/2014; 455(s 1–3):180–184.
[Show abstract][Hide abstract] ABSTRACT: This paper presents a demonstration experiment of ion cyclotron wall
conditioning (ICWC) on TEXTOR covering all ITER wall conditioning aims
and discusses the implications for ITER. O2/He-ICWC applied
to erode carbon co-deposits removed 6.6 × 1021 C-atoms
(39 pulses, 158 s cumulated discharge time). Large oxygen retention (71%
of injected oxygen) prevented subsequent ohmic discharge initiation.
Plasma operation was recovered by a 1h47 multi-pulse D2-ICWC
procedure including pumping time between pulses with duty cycle of 2
s/20 s, cleaning the vessel from oxygen impurities, followed by a 23 min
He-ICWC procedure (2 s/20 s), applied to desaturate the deuterium-loaded
walls. A stable ohmic discharge was established on the first attempt
right after the recovery procedure. The discharges showed improved
density control and only slightly increased oxygen characteristic
radiation levels (1-1.5 times). After the recovery procedure 36%
of the injected O-atoms remained retained in the vessel, derived from
mass spectrometry measurements. This amount is in the estimated range
for storage in remote areas obtained from surface analysis of locally
exposed samples. The removed amount of oxygen by D2 and
He-ICWC obtained from mass spectrometry corresponds to the retention in
plasma-wetted areas estimated by surface analysis. It is concluded that
most of the removed oxygen stems from plasma-wetted areas while shadowed
areas, e.g. behind poloidal limiters, may feature net retention of the
discharge gas. On ITER, designed with a shaped first wall, the ICWC
plasma-wetted area will approach the total surface area, reducing
consequently the retention in remote areas. A tentative extrapolation of
the carbon removal on TEXTOR to tritium removal from co-deposits on ITER
in the 39 × 4 s O2/He-ICWC discharges, including
pumping time between the RF pulses, corresponds on ITER to a tritium
removal in the order of the estimated retention per 400 s DT-burn
(140-500 mgT (Shimada and Pitts 2011 J. Nucl. Mater. 415
[Show abstract][Hide abstract] ABSTRACT: This paper focuses on a study of the principal operation aspects of standard ICRF heating antennas in the
ion cyclotron wall conditioning (ICWC) mode: (i) ability of the antenna to ignite the cleaning discharge
safely and reliably in different gases including those most likely to be used in ITER – He, H2, D2 and their
mixtures, (ii) the antenna capacity to couple a large fraction of the RF generator power (>50%) to low density (�1016–1018 m-�3) plasmas and (iii) the RF power absorption schemes aimed at improved RF plasma homogeneity and enhanced conditioning effect. The ICWC discharge optimization in terms of RF plasma wave excitation/absorption resulted in successful simulation of the conditioning scenarios for ITER operation at full field (JET) and half-field (TEXTOR, TORE SUPRA, ASDEX Upgrade).
Journal of Nuclear Materials 08/2013; 419:S1029–S1032. · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: This paper reports on the recent assessment of the Ion Cyclotron Wall Conditioning (ICWC) technique for isotopic ratio control, fuel removal and recovery after disruptions, which has been performed on TORE SUPRA, TEXTOR, ASDEX Upgrade and JET. ICWC discharges were produced using the standard ICRF heating antennas of each device, at different frequencies and toroidal fields, either in continuous or pulsed mode. Intrinsic ICWC discharge inhomogeneities could be partly compensated by applying a small vertical magnetic field, resulting in the vertical extension of the discharge in JET and TEXTOR. The conditioning efficiency was assessed from the flux of desorbed and retained species, measured by means of mass spectrometry. In Helium ICWC discharges, fuel removal rates between 1016D.m-2.s-1 to 3.1017D.m-2.s-1 were measured, with a linear dependence on the coupled RF power and on the He +
density. ICWC scenarios have been developed in D or H plasmas for isotopic exchange. The H (or D) outgassing was found to increase with the D (resp. H) partial pressure. In continuous mode, wall retention is on the average two to ten times higher than desorption
, due to the high reionization probability of desorbed species in ICWC discharges, where the electron density is about 1018m-3. Retention can be minimized in pulsed ICWC discharges without severely reducing outpumping. Pulsed He-ICWC discharges have been successfully used on TORE SUPRA to recover normal operation after disruptions,
when subsequent plasma initiation would not have been possible without conditioning.
Journal of Nuclear Materials 08/2013; 415:S1021–S1028. · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the understanding and validate the models of the in vessel build-up of the T inventory in ITER and future D–T devices. So far, research in these areas is largely supported by post-mortem analysis of wall tiles. However, access to samples will be very much restricted in the next-generation devices (such as ITER, JT-60SA, W7-X, etc) with actively cooled plasma-facing components (PFC) and increasing duty cycle.
[Show abstract][Hide abstract] ABSTRACT: This paper presents an analysis of the carbon–deuterium circulation and the resulting balance in Tore Supra over the period 2002–2007. Carbon balance combines the estimation of carbon gross erosion from spectroscopy, net erosion and deposition using confocal microscopy, lock-in thermography and SEM, and a measure of the amount of deposits collected in the vacuum chamber. Fuel retention is determined from post-mortem (PM) analyses and gas balance (GB) measurements. Special attention was paid to the deuterium outgassed during the nights and weekends of the experimental campaign (vessel under vacuum, Plasma Facing Components at 120 °C) and during vents (vessel at atmospheric pressure, PFCs at room temperature). It is shown that this outgassing is the main process reconciling the PM and GB estimations of fuel retention, closing the coupled carbon–deuterium balance. In particular, it explains why the deuterium concentration in deposits decreases with increasing depth.
Journal of Nuclear Materials 07/2013; 438(Supplement):S120-S125. · 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: Long term fuel retention experiments have been performed in JET with the ITER- Like Wall (JET-ILW) and compared with reference discharges in the preceding phase with carbon wall (JET-C) in ohmic, L and H-mode plasma scenarios. The long term fuel retention is evaluated through global gas balance with an accuracy of 1.2% for series of repetitive pulses (10–25) carried out over a full day of experiments. Compared to carbon wall, with the JET-ILW, for L mode, the retention exhibits also a drop from 1.26 × 1021 Ds−1 to 4–8 × 1019 Ds−1. Finally for Type III and type I ELMy H-mode, the retention decreases from 1.37 × 1021 Ds−1 to 7.2 × 1019 Ds−1 and from 1.7 × 1021 Ds−1 to 2.7 × 1020 Ds−1 respectively.The retention rates with the JET-ILW exhibit a decrease by a factor of about 10 compared to JET-C attributed to a reduction of carbon impurities and less fuel content in Be codeposition.
Journal of Nuclear Materials 07/2013; 438:S108–S113. · 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: Deuterium retention in Toughened, Fine-Grained Recrystallized W (TFGR
W-1.1 wt%TiC) was studied, compared to pure W. D implantation was
performed to a fluence of 1 × 1024 m‑2
at temperatures of 473-873 K, followed by TDS. It was found that D
retention in TFGR W is higher than in pure W at all irradiation
temperatures. Namely, at 673 K, D retention in TFGR W is six times
higher than pure W. TDS spectrum of TFGR W irradiated at 573 K has a
large peak around ˜700 K with small shoulder up to ˜1100 K.
In the case of D + He simultaneous irradiation, D retention is about 30%
lower than for pure D. In addition, plasma exposure experiment was also
conducted in TEXTOR, followed by NRA. Higher retention in TFGR W-1.1
wt%TiC could be attributed to high grain boundary diffusion (then
trapping deeper into the bulk) and formation of TiD2.
Journal of Nuclear Materials 07/2013; · 2.02 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The 1.1-1.5 mm wide gaps between tiles of the main toroidal belt limiter
in TEXTOR were utilized to study the long-term impurity deposition and
fuel retention in gaps. The tiles were exposed during a full tokamak
campaign of 9365 s of plasma to various discharge conditions and wall
conditioning, accumulating of up to 30 μm thick layers at the gap
entrance. It was found that (i) gaps trap impurities twice as efficient
as the top surface, (ii) the deposition in the toroidal gaps is twice as
high as in the poloidal, (iii) carbon deposition decays with a fall-off
length of about 0.7 mm towards the gap bottom, (iv) deposition on the
bottom is significantly higher than on the adjacent side walls of gaps,
and (v) the amount of deuterium scales with the amount of carbon with
D/C varying from 3% to 30% depending on the surface temperature.
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)