| 1. |
- Andersson Sundén, Erik, et al.
(författare)
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An assessment of nitrogen concentrations from spectroscopic measurements in the JET and ASDEX upgrade divertor
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 147-152
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Tidskriftsartikel (refereegranskat)abstract
- The impurity concentration in the tokamak divertor plasma is a necessary input for predictive scaling of divertor detachment, however direct measurements from existing tokamaks in different divertor plasma conditions are limited. To address this, we have applied a recently developed spectroscopic N II line ratio technique for measuring the N concentration in the divertor to a range of H-mode and L-mode plasma from the ASDEX Upgrade and JET tokamaks, respectively. The results from both devices show that as the power crossing the separatrix, P-sep, is increased under otherwise similar core conditions (e.g. density), a higher N concentration is required to achieve the same detachment state. For example, the N concentrations at the start of detachment increase from approximate to 2% to approximate to 9% as P-sep, is increased from approximate to 2.5 MW to approximate to 7 MW. These results tentatively agree with scaling law predictions (e.g. Goldston et al.) motivating a further study examining the parameters which affect the N concentration required to reach detachment. Finally, the N concentrations from spectroscopy and the ratio of D and N gas valve fluxes agree within experimental uncertainty only when the vessel surfaces are fully-loaded with N.
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| 2. |
- Bobkov, V, et al.
(författare)
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Impact of ICRF on the scrape-off layer and on plasma wall interactions : From present experiments to fusion reactor
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 131-140
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Tidskriftsartikel (refereegranskat)abstract
- Recent achievements in studies of the effects of ICRF (Ion Cyclotron Range of Frequencies) power on the SOL (Scrape-Off Layer) and PWI (Plasma Wall Interactions) in ASDEX Upgrade (AUG), Alcator C-Mod, and JET-ILW are reviewed. Capabilities to diagnose and model the effect of DC biasing and associated impurity production at active antennas and on magnetic field connections to antennas are described. The experiments show that ICRF near-fields can lead not only to E x B convection, but also to modifications of the SOL density, which for Alcator C-Mod are limited to a narrow region near antenna. On the other hand, the SOL density distribution along with impurity sources can be tailored using local gas injection in AUG and JET-ILW with a positive effect on reduction of impurity sources. The technique of RF image current cancellation at antenna limiters was successfully applied in AUG using the 3-strap AUG antenna and extended to the 4-strap Alcator C-Mod field-aligned antenna. Multiple observations confirmed the reduction of the impact of ICRF on the SOL and on total impurity production when the ratio of the power of the central straps to the total antenna power is in the range 0.6 < P-cen / P-total < 0.8. Near-field calculations indicate that this fairly robust technique can be applied to the ITER ICRF antenna, enabling the mode of operation with reduced PWI. On the contrary, for the A2 antenna in JET-ILW the technique is hindered by RF sheaths excited at the antenna septum. Thus, in order to reduce the effect of ICRF power on PWI in a future fusion reactor, the antenna design has to be optimized along with design of plasmafacing components.
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| 3. |
- Borodin, D., et al.
(författare)
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Improved ERO modelling of beryllium erosion at ITER upper first wall panel using JET-ILW and PISCES-B experience
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 510-515
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Tidskriftsartikel (refereegranskat)abstract
- ERO is a 3D Monte-Carlo impurity transport and plasma-surface interaction code. In 2011 it was applied for the ITER first wall (FW) life time predictions [1] (critical blanket module BM11). After that the same code was significantly improved during its application to existing fusion-relevant plasma devices: the tokamak JET equipped with an ITER-like wall and linear plasma device PISCES-B. This has allowed testing the sputtering data for beryllium (Be) and showing that the "ERO-min" fit based on the large (50%) deuterium (D) surface content is well suitable for plasma-wetted areas (D plasma). The improved procedure for calculating of the effective sputtering yields for each location along the plasma-facing surface using the recently developed semi-analytical sheath approach was validated. The re-evaluation of the effective yields for BM11 following the similar revisit of the JET data has indicated significant increase of erosion and motivated the current re-visit of ERO simulations.
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| 4. |
- Denis, J., et al.
(författare)
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Dynamic modelling of local fuel inventory and desorption in the whole tokamak vacuum vessel for auto-consistent plasma-wall interaction simulations
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 550-557
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Tidskriftsartikel (refereegranskat)abstract
- An extension of the SolEdge2D-EIRENE code package, named D-WEE, has been developed to add the dynamics of thermal desorption of hydrogen isotopes from the surface of plasma facing materials. To achieve this purpose, D-WEE models hydrogen isotopes implantation, transport and retention in those materials. Before launching autoconsistent simulation (with feedback of D-WEE on SolEdge2D-EIRENE), D-WEE has to be initialised to ensure a realistic wall behaviour in terms of dynamics (pumping or fuelling areas) and fuel content. A methodology based on modelling is introduced to perform such initialisation. A synthetic plasma pulse is built from consecutive SolEdge2D-EIRENE simulations. This synthetic pulse is used as a plasma background for the D-WEE module. A sequence of plasma pulses is simulated with D-WEE to model a tokamak operation. This simulation enables to extract at a desired time during a pulse the local fuel inventory and the local desorption flux density which could be used as initial condition for coupled plasma-wall simulations. To assess the relevance of the dynamic retention behaviour obtained in the simulation, a confrontation to post-pulse experimental pressure measurement is performed. Such confrontation reveals a qualitative agreement between the temporal pressure drop obtained in the simulation and the one observed experimentally. The simulated dynamic retention during the consecutive pulses is also studied.
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| 5. |
- Hatano, Y., et al.
(författare)
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Tritium distributions on W-coated divertor tiles used in the third JET ITER-like wall campaign
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 258-261
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Tidskriftsartikel (refereegranskat)abstract
- Tritium (T) distributions on tungsten (W)-coated plasma-facing tiles used in the third ITER-like wall campaign (2015-2016) of the Joint European Torus (JET) were examined by means of an imaging plate technique and beta-ray induced x-ray spectrometry, and they were compared with the distributions after the second (2013-2014) campaign. Strong enrichment of T in beryllium (Be) deposition layers was observed after the second campaign. In contrast, T distributions after the third campaign was more uniform though Be deposition layers were visually recognized. The one of the possible explanations is enhanced desorption of T from Be deposition layers due to higher tile temperatures caused by higher energy input in the third campaign.
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| 6. |
- Heinola, K., et al.
(författare)
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Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 19, s. 397-402
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Tidskriftsartikel (refereegranskat)abstract
- Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate.
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| 7. |
- Huber, A., et al.
(författare)
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Determination of tungsten sources in the JET-ILW divertor by spectroscopic imaging in the presence of a strong plasma continuum
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 18, s. 118-124
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Tidskriftsartikel (refereegranskat)abstract
- The identification of the sources of atomic tungsten and the measurement of their radiation distribution in front of all plasma-facing components has been performed in JET with the help of two digital cameras with the same two-dimensional view, equipped with interference filters of different bandwidths centred on the W I (400.88 nm) emission line. A new algorithm for the subtraction of the continuum radiation was successfully developed and is now used to evaluate the W erosion even in the inner divertor region where the strong recombination emission is dominating over the tungsten emission. Analysis of W sputtering and W redistribution in the divertor by video imaging spectroscopy with high spatial resolution for three different magnetic configurations was performed. A strong variation of the emission of the neutral tungsten in toroidal direction and corresponding W erosion has been observed. It correlates strongly with the wetted area with a maximal W erosion at the edge of the divertor tile.
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| 8. |
- Kirschner, A., et al.
(författare)
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Modelling of tungsten erosion and deposition in the divertor of JET-ILW in comparison to experimental findings
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 239-244
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Tidskriftsartikel (refereegranskat)abstract
- The erosion, transport and deposition of tungsten in the outer divertor of JET-ILW has been studied for an H-Mode discharge with low frequency ELMs. For this specific case with an inter-ELM electron temperature at the strike point of about 20 eV, tungsten sputtering between ELMs is almost exclusively due to beryllium impurity and self-sputtering. However, during ELMs tungsten sputtering due to deuterium becomes important and even dominates. The amount of simulated local deposition of tungsten relative to the amount of sputtered tungsten in between ELMs is very high and reaches values of 99% for an electron density of 5E13 cm(-3) at the strike point and electron temperatures between 10 and 30 eV. Smaller deposition values are simulated with reduced electron density. The direction of the B-field significantly influences the local deposition and leads to a reduction if the E x B drift directs towards the scrape-off-layer. Also, the thermal force can reduce the tungsten deposition, however, an ion temperature gradient of about 0.1 eV/mm or larger is needed for a significant effect. The tungsten deposition simulated during ELMs reaches values of about 98% assuming ELM parameters according to free-streaming model. The measured WI emission profiles in between and within ELMs have been reproduced by the simulation. The contribution to the overall net tungsten erosion during ELMs is about 5 times larger than the one in between ELMs for the studied case. However, this is due to the rather low electron temperature in between ELMs, which leads to deuterium impact energies below the sputtering threshold for tungsten.
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| 9. |
- Krivska, A., et al.
(författare)
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RF sheath modeling of experimentally observed plasma surface interactions with the JET ITER-Like Antenna
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 19, s. 324-329
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Tidskriftsartikel (refereegranskat)abstract
- Waves in the Ion Cyclotron Range of Frequencies (ICRF) enhance local Plasma-Surface Interactions (PSI) near the wave launchers and magnetically-connected objects via Radio-Frequency (RF) sheath rectification. ITER will use 20MW of ICRF power over long pulses, questioning the long-term impact of RF-enhanced localized erosion on the lifetime of its Beryllium (Be) wall. Recent dedicated ICRF-heated L-mode discharges documented this process on JET for different types of ICRF antennas. Using visible spectroscopy in JET ICRF-heated L-mode discharges, poloidally-localized regions of enhanced (by similar to 2-4x) Be I and Be II light emission were observed on two outboard limiters magnetically connected to the bottom of the active ITER-Like Antenna (ILA). The observed RF-PSI induced by the ILA was qualitatively comparable to that induced by the JET standard, type-A2 antennas, for similar strap toroidal phasing and connection geometries. The Be II line emission was found more intense when powering the bottom half of the ILA rather than its top half. Conversely, more pronounced SOL density modifications were observed with only top array operation, on field lines connected to the top half of the ILA. So far the near-field modeling of the ILA with antenna code TOPICA (Torino Polytechnic Ion Cyclotron Antenna), using curved antenna model, was partially able to reproduce qualitatively the observed phenomena. A quantitative discrepancy persisted between the observed Be source amplification and the calculated, corresponding increases in E-// field at the magnetically connected locations to the ILA when changing from only top to only bottom half antenna operation. This paper revisits these current drive phased and half-ILA powered cases using for the new simulations flat model of the ILA and more realistic antenna feeding to calculate the E-// field maps with TOPICA code. Further, the Self-consistent Sheaths and Waves for Ion Cyclotron Heating Slow Wave (SSWICH-SW) code, which couples slow wave evanescence with DC Scrape-Off Layer (SOL) biasing, is used to estimate the poloidal distribution of rectified RF-sheath Direct Current (DC) potential V-DC in the private SOL between the ILA poloidal limiters. The approach so far was limited to correlating the observed, enhanced emission regions at the remote limiters to the antenna near-electric fields, as calculated by TOPICA. The present approach includes also a model for the rectification of these near-fields in the private SOL of the ILA. With the improved approach, when comparing only top and only bottom half antenna feeding, we obtained good qualitative correlation between all experimental measurements and the calculated local variations in the E-// field and V-DC potential.
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| 10. |
- Likonen, J., et al.
(författare)
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Investigation of deuterium trapping and release in the JET divertor during the third ILW campaign using TDS
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 300-306
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Tidskriftsartikel (refereegranskat)abstract
- Selected set of samples from JET ITER-Like Wall (JET-ILW) divertor tiles exposed in 2015-2016 has been analysed using Thermal Desorption Spectrometry (TDS). The deuterium (D) amounts obtained with TDS were compared with Nuclear Reaction Analysis (NRA). The highest amount of D was found on the top part of inner divertor which has regions with the thickest deposited layers as for divertor tiles removed in 2014. This area resides deep in the scrape-off layer and plasma configurations for the second (ILW-2, 2013-2014) and the third (ILW-3, 2015-2016) JET-ILW campaigns were similar. Agreement between TDS and NRA is good on the apron of Tile 1 and on the upper vertical region whereas on the lower vertical region of Tile 1 the NRA results are clearly smaller than the TDS results. Inner divertor Tile 3 has somewhat less D than Tiles 0 and 1, and the D amount decreases towards the lower part of the tile. The D retention at the divertor inner and outer corner regions is not symmetric as there is more D retention poloidally at the inner than at the outer divertor corner. In most cases the TDS spectra for the ILW-3 samples are different from the corresponding ILW-2 spectra because HD and D-2 release occurs at higher temperatures than from the ILW-2 samples indicating that the low energy traps have been emptied during the plasma operations and that D is either in the energetically deep traps or located deeper in the sample.
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| 11. |
- Likonen, J., et al.
(författare)
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Investigation of deuterium trapping and release in the JET ITER-like wall divertor using TDS and TMAP
- 2019
-
Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 166-178
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Tidskriftsartikel (refereegranskat)abstract
- Selected set of samples from JET ITER-Like Wall (JET-ILW) divertor tiles exposed both in 2013-2014 and 2011-2014 has been analysed using Thermal Desorption Spectrometry (TDS). The deuterium (D) amounts obtained with TDS were compared with Ion Beam Analysis (IBA) and Secondary Ion Mass Spectrometry (SIMS) data. The highest amount of D was found on the top part of inner divertor which has regions with the thickest deposited layers. This area resides deep in the scrape-off layer. Changes in plasma configurations between the first (2011-2012) and the second (2013-2014) JET-ILW campaign altered the material migration towards the inner and the outer divertor corner increasing the amount of deposition in the shadowed areas of the divertor base tiles. D retention on the outer divertor tiles is clearly smaller than on the inner divertor tiles. Experimental TDS spectra for samples from the top part of inner divertor and from the outer strike point region were modelled using TMAP program. Experimental deuterium profiles obtained with SIMS have been used and the detrapping and the activation energies have been adjusted. Analysis of the results of the TMAP simulations enabled to determine the nature of traps in different samples.
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| 12. |
- Makepeace, C., et al.
(författare)
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The effect of beryllium oxide on retention in JET ITER-like wall tiles
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 346-351
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Tidskriftsartikel (refereegranskat)abstract
- Preliminary results investigating the microstructure, bonding and effect of beryllium oxide formation on retention in the JET ITER-like wall beryllium tiles, are presented. The tiles have been investigated by several techniques: Scanning Electron Microscopy (SEM) equipped with Energy Dispersive X-ray (EDX), Transmission Electron microscopy (TEM) equipped with EDX and Electron Energy Loss Spectroscopy (EELS), Raman Spectroscopy and Thermal Desorption Spectroscopy (TDS). This paper focuses on results from melted materials of the dump plate tiles in JET. From our results and the literature, it is concluded, beryllium can form micron deep oxide islands contrary to the nanometric oxides predicted under vacuum conditions. The deepest oxides analyzed were up to 2-micron thicknesses. The beryllium Deuteroxide (BeOxDy) bond was found with Raman Spectroscopy. Application of EELS confirmed the oxide presence and stoichiometry. Literature suggests these oxides form at temperatures greater than 700 degrees C where self-diffusion of beryllium ions through the surface oxide layer can occur. Further oxidation is made possible between oxygen plasma impurities and the beryllium ions now present at the wall surface. Under Ultra High Vacuum (UHV) nanometric Beryllium oxide layers are formed and passivate at room temperature. After continual cyclic heating (to the point of melt formation) in the presence of oxygen impurities from the plasma, oxide growth to the levels seen experimentally (approximately two microns) is proposed. This retention mechanism is not considered to contribute dramatically to overall retention in JET, due to low levels of melt formation. However, this mechanism, thought the result of operation environment and melt formation, could be of wider concern to ITER, dependent on wall temperatures.
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| 13. |
- Moon, Sunwoo, et al.
(författare)
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First mirror test in JET for ITER : Complete overview after three ILW campaigns
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 59-66
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Tidskriftsartikel (refereegranskat)abstract
- The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015-2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011-2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20-80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5 x 10(17) cm(-2), W is up to 4.3 x 10(17) cm(-2), while the content of Ni is the greatest in the outer divertor: 3.8 x 10(17) cm(-2). Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 x 10(16) cm(-2). The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s(-1)) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred.
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| 14. |
- Pajuste, Elina, et al.
(författare)
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Comparison of the structure of the plasma-facing surface and tritium accumulation in beryllium tiles from JET ILW campaigns 2011-2012 and 2013-2014
- 2019
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Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 19, s. 131-136
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Tidskriftsartikel (refereegranskat)abstract
- In this study, beryllium tiles from Joint European Torus (JET) vacuum vessel wall were analysed and compared regarding their position in the vacuum vessel and differences in the exploitation conditions during two campaigns of ITER-Like-Wall (ILW) in 2011-2012 (ILW1) and 2013-2014 (ILW2) Tritium content in beryllium samples were assessed. Two methods were used to measure tritium content in the samples - dissolution under controlled conditions and tritium thermal desorption. Prior to desorption and dissolution experiments, scanning electron microscopy and energy dispersive x-ray spectroscopy were used to study structure and chemical composition of plasma-facing-surfaces of the beryllium samples. Experimental results revealed that tritium content in the samples is in range of 2.10(11)-2.10(13) tritium atoms per square centimetre of the surface area with its highest content in the samples from the outer wall of the vacuum vessel (up to 1.9.10(13) atoms/cm(2) in ILW1 campaign and 2.4.10(13) atoms/cm(2) in ILW2). The lowest content of tritium was found in the upper part of the vacuum vessel (2.0.10(12) atoms/cm(2) and 2.0.10(11) atoms/cm(2) in ILW1 and ILW2, respectively). Results obtained from scanning electron microscopy has shown that surface morphology is different within single tile, however if to compare two campaigns main tendencies remains similar.
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| 15. |
- Romazanov, J., et al.
(författare)
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Beryllium global erosion and deposition at JET-ILW simulated with ERO2.0
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 331-338
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Tidskriftsartikel (refereegranskat)abstract
- The recently developed Monte-Carlo code ERO2.0 is applied to the modelling of limited and diverted discharges at JET with the ITER-like wall (ILW). The global beryllium (Be) erosion and deposition is simulated and compared to experimental results from passive spectroscopy. For the limiter configuration, it is demonstrated that Be self-sputtering is an important contributor (at least 35%) to the Be erosion. Taking this contribution into account, the ERO2.0 modelling confirms previous evidence that high deuterium (D) surface concentrations of up to similar to 50% atomic fraction provide a reasonable estimate of Be erosion in plasma-wetted areas. For the divertor configuration, it is shown that drifts can have a high impact on the scrape-off layer plasma flows, which in turn affect global Be transport by entrainment and lead to increased migration into the inner divertor. The modelling of the effective erosion yield for different operational phases (ohmic, L- and H-mode) agrees with experimental values within a factor of two, and confirms that the effective erosion yield decreases with increasing heating power and confinement.
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| 16. |
- Tolias, Panagiotis, 1984-, et al.
(författare)
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The adhesion of tungsten dust on plasma-exposed tungsten surfaces
- 2019
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Ingår i: Nuclear Materials and Energy. - : ELSEVIER SCIENCE BV. - 2352-1791. ; 18, s. 18-22
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Tidskriftsartikel (refereegranskat)abstract
- The adhesion of tungsten dust is measured on plasma-exposed and non-exposed tungsten substrates with the electrostatic detachment method. Tungsten substrates of comparable surface roughness have been exposed to the deuterium plasmas of the GyM linear device and the argon plasmas of rf glow discharges under conditions which invariably modify the surface composition due to physical sputtering. The adhesion has been systematically characterized for different spherical nearly monodisperse dust populations. Independent of the dust size, an approximate 50% post-exposure reduction of the average and spread of the adhesive force has been consistently observed and attributed to surface chemistry modifications.
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| 17. |
- Vignitchouk, Ladislas, et al.
(författare)
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Accumulation of beryllium dust in ITER diagnostic ports after off-normal events
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 20
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Tidskriftsartikel (refereegranskat)abstract
- Dust transport simulations are used to predict the effect of diagnostic ports on the in-vessel inventory of solid beryllium particles produced during mitigated disruptions in ITER. Beryllium dust is assumed to originate from the re-solidification of liquid droplets, which are initially ejected during transient first-wall melting events and subsequently interact with the disrupting plasma. The trajectories of droplets launched with various initial conditions, as well as the time evolution of their temperature and mass, are simulated until either complete vaporization or immobilization upon undergoing a sticking or splashing impact with the wall is realized. The results indicate that approximately 10% of the dust mass in the vessel can be expected to reside inside ports, in particular those located in the equatorial plane or in the lower outboard first wall.
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| 18. |
- Weckmann, Armin, et al.
(författare)
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Physics affecting heavy impurity migration in tokamaks : Benchmarking test-ion code ASCOT against TEXTOR tracer experiment
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier Ltd. - 2352-1791. ; 19, s. 307-315
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Tidskriftsartikel (refereegranskat)abstract
- Erosion, transport and deposition of wall impurities are major concerns in future magnetic fusion devices, both from the perspective of the fusion plasma and the machine wall. An extensive study on molybdenum transport and deposition performed in the TEXTOR tokamak yielded a detailed deposition map that is ideal for benchmark deposition studies. A qualitative benchmark is attempted in this article with the ASCOT code. We set up a full 3D model of the TEXTOR tokamak and studied the influence of different physical mechanisms and their strengths on molybdenum deposition patterns on the simulated plasma-facing components: atomic processes, Coulomb collisions, scrape-off layer (SOL) profiles, source distribution, marker starting energy, radial electric field strength, SOL flow and toroidal plasma rotation. The outcome comprises 13 simulations, each with 100,000 markers. The findings are: • Toroidal plasma movement, either within the LCFS or as SOL flow, is negligible. • SOL profile and marker starting energy have modest impact on deposition. • Source distribution has a large impact in combination with radial electric field profiles. • The E⇀×B⇀ drift has the highest impact on the deposition profiles.
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| 19. |
- Widdowson, A., et al.
(författare)
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Deposition of impurity metals during campaigns with the JET ITER-like Wall
- 2019
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Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 218-224
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Tidskriftsartikel (refereegranskat)abstract
- Post mortem analysis shows that mid and high atomic number metallic impurities are present in deposits on JET plasma facing components with the highest amount of Ni and W, and therefore the largest sink, being found at the top of the inner divertor. Sources are defined as "continuous" or "specific", in that "continuous" sources arise from ongoing erosion from plasma facing surfaces and "specific" are linked with specific events which decrease over time until they no longer act as a source. This contribution evaluates the sinks and estimates sources, and the balance gives an indication of the dominating processes. Charge exchange neutral erosion is found to be the main source of nickel, whereas erosion of divertor plasma facing components is the main source of tungsten. Specific sources are shown to have little influence over the global mid- and high-Z impurity concentrations in deposits.
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| 20. |
- Zhou, Yushan, et al.
(författare)
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The effect of gyration on the deposition of beryllium and deuterium at rough surface on the divertor tiles with ITER-like-wall in JET
- 2019
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Ingår i: Nuclear Materials and Energy. - : ELSEVIER SCIENCE BV. - 2352-1791. ; 19, s. 155-160
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Tidskriftsartikel (refereegranskat)abstract
- Experimental results from microbeam analyses on divertor tiles, through 2011-2012 and 2013-2014 operations in JET-ILW, showed that the deuterium retention and beryllium impurity deposition were non-uniform on the rough surface in micro meter scale. Frequently, Be and D were accumulated within pits, cracks and valleys in the range of similar to 10 mu m to similar to 100 mu m and selectively at side slopes within larger pits. In the deposited layers, it was also observed that cylindrical structures grow in a specific direction with respect to the magnetic field. In this work, a sample from the divertor in JET-ILW following 2013-2014 operation was analysed. The surface topography was measured with the focus stacking technique on photos from optical microscope and elemental distribution maps were obtained by micro-ion-beam analysis. In order to figure out the effect of ion gyro motion on Be and D deposition on the real divertor topography, a modelling of ion trajectories in plasma boundary was performed. Ion beam analysis data suggested that D accumulated on higher areas on the sample, while Be stayed on lower positions. The gyration of ions in edge plasma and corresponding impact positions on the sample surface could explain some IBA results qualitatively.
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| 21. |
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