[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. · 1.21 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. · 1.21 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: 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; · 1.21 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. · 1.21 Impact Factor
[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. · 1.21 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. · 1.21 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: To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be up to 30% higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be re-established only when using gas puff levels of a few 1021 es−1. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at βN ≈ 3, hybrids) have been achieved with W concentrations well below the maximum acceptable level.
Physics of Plasmas 05/2013; 20(5). · 2.38 Impact Factor
[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: Ion cyclotron wall conditioning (ICWC) discharges, in pulsed-mode operation, were
carried out in the limiter tokamak TEXTOR to explore safe operational regimes for the experimental
parameters for possible ICWC-discharge cleaning in International Thermonuclear Experimental
Reactor (ITER) at half field. Antenna coupling properties obtained during the ion cyclotron range
of frequencies (ICRF) wall conditioning experiments performed in helium–hydrogen mixture in
TEXTOR were analysed in relation to the obtained ICWC-plasma characterization results. Satisfactory
antenna coupling in the mode conversion scenario along with reproducible generation of
ICRF plasmas for wall conditioning, were achieved by coupling radio frequency (RF) power from
one or two ICRF antennas. The plasma breakdown results obtained in the TEXTOR tokamak
have been compared with the predictions of a zero-dimensional RF plasma production model. The
present study of ICWC emphasizes the beneficial effect of application of an additional (along with
toroidal magnetic field) stationary vertical (BV � BT) or oscillating poloidal magnetic field (BP �
BT) on antenna coupling and relevant plasma parameters.
[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; · 1.21 Impact Factor
[show abstract][hide abstract] ABSTRACT: Results of a new dedicated experiment addressing the problem of impurity deposition at the bottom in gaps are presented along with modelling. A test limiter with an isolated gap was exposed to the scrape-off layer plasma in TEXTOR. The exposure was accompanied by injection of 13C-marked methane in the vicinity of the gap. Deposition at the bottom of the gap was monitored in situ with Quartz-Microbalance diagnostics. The 13C deposition efficiency of about 2.6 × 10−5 was measured. Post mortem analysis of resulting deposited layers performed with SIMS and EPMA techniques yields about a factor 2 smaller value corresponding to approximately 10% contribution of the gap bottom to the total 13C deposition in the gap. This measured contribution is effectively much smaller than observed earlier in TEXTOR, taking the difference in geometry into account, and is in reasonable agreement with modelling performed with ERO and 3D-GAPS codes.
Journal of Nuclear Materials 01/2013; · 1.21 Impact Factor
[show abstract][hide abstract] ABSTRACT: Beryllium (Be) erosion has been studied in dedicated limiter discharges in JET with the recently installed ITER-like wall . Passive spectroscopy  in the vicinity of the solid Be limiter is used for measurement of the Be physical sputtering as the main erosion mechanism. To consider the 3D configuration of plasma parameters and the electromagnetic field, the actual limiter shape and the local transport affecting the fraction of Be coming into the observation volume, a detailed modelling with the ERO code has been applied to interpret the experimental data. The observed dependence of BeI and BeII line intensities on plasma parameters during the density scan and various line ratios are used to validate the model and the underlying data including the recently introduced assumptions for Be physical sputtering, the very same which were used before for ITER predictive modelling .
Journal of Nuclear Materials 01/2013; · 1.21 Impact Factor
[show abstract][hide abstract] ABSTRACT: Fusion Engineering and Design j o u r n a l h o m e p a g e : w w w . e l s e v i e r . c o m / l o c a t e / f u s e n g d e s a b s t r a c t This paper focuses on encouraging results obtained on the characterization of RF produced plasmas during pulsed-mode wall conditioning discharges in ion cyclotron resonance frequency (ICRF) regime in the lim-iter tokamak TEXTOR. Recent Ion Cyclotron Wall Conditioning (ICWC) experiment carried out in TEXTOR tokamak, lead to the identification of various dependences of the antenna-plasma coupling efficiency on the plasma parameters for possible ICWC-discharge cleaning in ITER at half field. Our ICWC experiments emphasize on (i) study of antenna coupling during the mode conversion scenario, (ii) reproducible gen-eration of ICRF plasmas for wall conditioning, by coupling RF power from one or two ICRF antennas and (iii) effect of application of an additional (along with toroidal magnetic field) stationary vertical (B V B T) or oscillating poloidal magnetic field (B p B T) on antenna coupling and relevant plasma parameters.
Fusion Engineering and Design 01/2013; 88:51-56. · 0.84 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: 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: 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: Laser Induced Breakdown Spectroscopy (LIBS) is a potential candidate to
monitor the layer composition and fuel retention during and after plasma
shots on specific locations of the main chamber and divertor of ITER.
This method is being investigated in a cooperative research programme on
plasma devices such as TEXTOR, FTU, MAGNUM-PSI and in other various
laboratorial experiments. In this paper LIBS results from targets of
D-H-rich carbon films and mixed W-Al-C deposits on bulk tungsten
substrates are reported (simulating ITER-like deposits with Al as proxy
for Be). Two independent methods, one to determine the relative
elemental composition and the other the absolute contents of the target
based on the experimental LIBS signals are proposed. The results show
that LIBS has the capability to provide the relative concentrations of
the elements on the deposited layer when the experimental conditions on
the targets surface are identical to the calibration samples.