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1.
  • Andersson Sundén, Erik, et al. (författare)
  • An assessment of nitrogen concentrations from spectroscopic measurements in the JET and ASDEX upgrade divertor
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 147-152
  • 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.
  • Ashikawa, N., et al. (författare)
  • Determination of retained tritium from ILW dust particles in JET
  • 2020
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 22
  • Tidskriftsartikel (refereegranskat)abstract
    • Quantitative tritium inventory in dust particles from campaigns in the JET tokamak with the carbon wall (2007–2009) and the ITER-like wall (ILW 2011–2012) were determined by the liquid scintillation counter and the full combustion method. A feature of this full combustion method is that dust particles were covered by a tin (Sn) which reached 2100 K during combustion under oxygen flow. The specific tritium inventory for samples from JET with carbon and with metal walls was measured and found to be similar. However, the total tritium inventory in dust particles from the ILW experiment was significantly smaller in comparison to the carbon wall due to the lower amount of dust particles generated in the presence of metal walls.
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3.
  • Balbinot, L., et al. (författare)
  • Multi-code estimation of DTT edge transport parameters
  • 2023
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 34
  • Tidskriftsartikel (refereegranskat)abstract
    • The main goal of the Divertor Tokamak Test facility (DTT) is to operate with a high value of power-exhaust-relevant parameter Psoz/R in plasma scenarios similar to those foreseen for the Demonstration Fusion Power Plant (DEMO) in terms of low collisionality and neutral opacity. For these unique characteristics, accurate modelling of the principal scenario is necessary for machine designing. In edge numerical codes, cross-field transport profiles have a high impact on modelling results. This work aims at providing a coherent set of transport parameters for DTT full-power (FP) single-null (SN) scenario edge modelling. To evaluate such parameters for DTT, a transport analysis on the current machine has been performed using SOLEDGE2D-EIRENE and SOLPS-ITER. The transport parameters to be used in the simulations of the DTT single-null scenario were selected using two complementary methods. The first is the modelling of JET and Alcator C-Mod (C-Mod) with SOLEDGE2D-EIRENE and SOLPS-ITER, validating transport parameters by comparing modelling results to experimental data from pulses which are considered DTT-relevant. JET pulses were selected with the highest auxiliary input power (from 29 to 33 MW), plasma current and toroidal field to better match DTT parameters; nitrogen and neon seeded pulses were selected to check possible seeding material dependencies. The considered C-Mod pulse better matches DTT plasma density and neutral opacity. Transport parameters are then scaled to DTT according to scaling laws. The second method derives the transport parameters by tuning their values inside the DTT separatrix to reproduce the pedestal profiles predicted by the EPED model via the Europed code and applied in DTT. The derived set of DTT transport parameters is consistent with the results obtained by modelling present machines, reproduces the expected heat flux decay length in detached conditions and, inside the separatrix, reproduces the predicted pedestal using transport parameters which are coherent with what is predicted by the quasi-linear turbulent model QuaLiKiz.
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4.
  • Bernert, M., et al. (författare)
  • Power exhaust by SOL and pedestal radiation at ASDEX Upgrade and JET
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 111-118
  • Tidskriftsartikel (refereegranskat)abstract
    • Future fusion reactors require a safe, steady state divertor operation. A possible solution for the power exhaust challenge is the detached divertor operation in scenarios with high radiated power fractions. The radiation can be increased by seeding impurities, such as N for dominant scrape-off-layer radiation, Ne or Ar for SOL and pedestal radiation and Kr for dominant core radiation. Recent experiments on two of the all-metal tokamaks, ASDEX Upgrade (AUG) and JET, demonstrate operation with high radiated power fractions and a fully-detached divertor by N, Ne or Kr seeding with a conventional divertor in a vertical target geometry. For both devices similar observations can be made. In the scenarios with the highest radiated power fraction, the dominant radiation originates from the confined region, in the case of N and Ne seeding concentrated in a region close to the X-point. Applying these seed impurities for highly radiative scenarios impacts local plasma parameters and alters the impurity transport in the pedestal region. Thus, plasma confinement and stability can be affected. A proper understanding of the effects by these impurities is required in order to predict the applicability of such scenarios for future devices.
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5.
  • Bobkov, V, et al. (författare)
  • Impact of ICRF on the scrape-off layer and on plasma wall interactions : From present experiments to fusion reactor
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 131-140
  • 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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6.
  • Bobkov, V., et al. (författare)
  • Progress in reducing ICRF-specific impurity release in ASDEX upgrade and JET
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 12, s. 1194-1198
  • Tidskriftsartikel (refereegranskat)abstract
    • Use of new 3-strap ICRF antennas with all-tungsten (W) limiters in ASDEX Upgrade results in a reduction of the W sources at the antenna limiters and of the W content in the confined plasma by at least a factor of 2 compared to the W-limiter 2-strap antennas used in the past. The reduction is observed with a broad range of plasma shapes. In multiple locations of antenna frame, the limiter W source has a minimum when RF image currents are decreased by cancellation of the RF current contributions of the central and the outer straps. In JET with ITER-like wall, ITER-like antenna produces about 20% less of main chamber radiation and of W content compared to the old A2 antennas. However the effect of the A2 antennas on W content is scattered depending on which antennas are powered. Experiments in JET with trace nitrogen (N-2) injection show that a presence of active ICRF antenna close to the midplane injection valve has little effect on the core N content, both in dipole and in -90 degrees phasing. This indicates that the effect of ICRF on impurity transport across the scape-off-layer is small in JET compared to the dominant effect on impurity sources leading to increased impurity levels during ICRF operation.
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7.
  • Borodin, D., et al. (författare)
  • Improved ERO modelling for spectroscopy of physically and chemically assisted eroded beryllium from the JET-ILW
  • 2016
  • Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 9, s. 604-609
  • Tidskriftsartikel (refereegranskat)abstract
    • Physical and chemical assisted physical sputtering were characterised by the Be I and Be II line and BeD band emission in the observation chord measuring the sightline integrated emission in front of the inner beryllium limiter at the torus midplane. The 3D local transport and plasma-surface interaction Monte-Carlo modelling (ERO code [18]) is a key for the interpretation of the observations in the vicinity of the shaped solid Be limiter. The plasma parameter variation (density scan) in limiter regime has provided a useful material for the simulation benchmark. The improved background plasma parameters input, the new analytical expression for particle tracking in the sheath region and implementation of the BeD release into ERO has helped to clarify some deviations between modelling and experiments encountered in the previous studies [4,5]. Reproducing the observations provides additional confidence in our 'ERO-min' fit for the physical sputtering yields for the plasma-wetted areas based on simulated data.
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8.
  • Borodin, D., et al. (författare)
  • Improved ERO modelling of beryllium erosion at ITER upper first wall panel using JET-ILW and PISCES-B experience
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 510-515
  • 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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9.
  • Borodkina, I., et al. (författare)
  • An analytical expression for ion velocities at the wall including the sheath electric field and surface biasing for erosion modeling at JET ILW
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 341-345
  • Tidskriftsartikel (refereegranskat)abstract
    • For simulation of plasma-facing component erosion in fusion experiments, an analytical expression for the ion velocity just before the surface impact including the local electric field and an optional surface biasing effect is suggested. Energy and angular impact distributions and the resulting effective sputtering yields were produced for several experimental scenarios at JET ILW mostly involving PFCs exposed to an oblique magnetic field. The analytic solution has been applied as an improvement to earlier ERO modelling of localized, Be outer limiter, RF-enhanced erosion, modulated by toggling of a remote, however magnetically connected ICRH antenna. The effective W sputtering yields due to D and Be ion impact in Type-I and Type-III ELMs and inter-ELM conditions were also estimated using the analytical approach and benchmarked by spectroscopy. The intra-ELM W sputtering flux increases almost 10 times in comparison to the inter-ELM flux.
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10.
  • Bykov, I., et al. (författare)
  • Modification of adhered dust on plasma-facing surfaces due to exposure to ELMy H-mode plasma in DIII-D
  • 2017
  • Ingår i: NUCLEAR MATERIALS AND ENERGY. - : Elsevier BV. - 2352-1791. ; 12, s. 379-385
  • Tidskriftsartikel (refereegranskat)abstract
    • Transient heat load tests have been conducted in the lower divertor of DIII-D using DiMES manipulator in order to study the behavior of dust on tungsten Plasma Facing Components (PFCs) during ELMy H-mode discharges. Samples with pre- adhered, pre- characterized dust have been exposed at the outer strike point (OSP) in a series of discharges with varied intra-(inter-) ELM heat fluxes. We used C dust because of its high sublimation temperature and non-metal properties. Al dust as a surrogate for Be and W dust were employed as relevant to that in the ITER divertor. The poor initial thermal contact between the substrate and the particles led to overheating, sublimation and shrinking of the carbon dust, and wetting induced coagulation of Al dust. Little modification of the W dust was observed. An enhanced surface adhesion and improvement of the thermal contact of C and Al dust were the result of exposure. A post mortem "adhesive tape" sampling showed that 70% of Al, <5% of W and C particles could not be removed from the surface owing to the improved adhesion. Al and C but not W particles that could be lifted had W inclusions indicating damage to the substrate. This suggests that non destructive methods may be inefficient for removal of dust in ITER.
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11.
  • Catarino, N., et al. (författare)
  • Assessment of erosion, deposition and fuel retention in the JET-ILW divertor from ion beam analysis data
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 12, s. 559-563
  • Tidskriftsartikel (refereegranskat)abstract
    • Post-mortem analyses of individual components provide relevant information on plasma-surface interactions like tungsten erosion, beryllium deposition and plasma fuel retention with divertor tiles via implantation or co-deposition. Ion Beam techniques are ideal tools for such purposes and have been extensively used for post-mortem analyses of selected tiles from JET following each campaign. In this contribution results from tiles removed from the JET ITER-Like Wall (JET-ILW) divertor following the 2013-2014 campaign are presented. The results summarize erosion, deposition and fuel retention along the poloidal cross section of the divertor surface and provide data for comparison with the first JET-ILW campaign, showing a similar pattern of material migration with the exception of Tile 6 where the strike point time on the tile was similar to 4 times longer in 2013-2014 than in 2011-2012, which is likely to account for more material migration to this region. The W deposition on top of the Mo marker coating of Tile 4 shows that the enrichment takes place at the strike point location. (C) 2016 The Authors. Published by Elsevier Ltd.
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12.
  • Chankina, A. V., et al. (författare)
  • Possible influence of near SOL plasma on the H-mode power threshold
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 273-277
  • Tidskriftsartikel (refereegranskat)abstract
    • A strong effect of divertor configuration on the threshold power for the L-H transition (P-LH) was observed in recent JET experiments in the new ITER-like Wall (ILW) [1-3]. Following a series of EDGE2D-EIRENE code simulations with Be impurity and drifts a possible mechanism for the P-LH variation with the divertor geometry is proposed. Both experiment and code simulations show that in the configuration with lower neutral recycling near the outer strike point (OSP), electron temperature (T-e) peaks near the OSP prior to the L-H transition, while in the configuration with higher OSP recycling T-e peaks further out in the scrape-offlayer (SOL) and the plasma stays in the L-mode at the same input power. Code results show large positive radial electric field (E-r) in the near SOL under lower recycling conditions leading to a large E x B shear across the separatrix which may trigger earlier (at lower input power) edge turbulence suppression and lower P-LH. Suppressed T-e's at OSP in configurations with strike points on vertical targets (VT) were observed earlier and explained by a geometrical effect of neutral recycling near this particular position, whereas in configurations with strike points on horizontal targets (HT) the OSP appears to be more open for neutrals (see e.g. review paper [4]).
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13.
  • Coburn, J., et al. (författare)
  • Reassessing energy deposition for the ITER 5 MA vertical displacement event with an improved DINA model
  • 2021
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 28
  • Tidskriftsartikel (refereegranskat)abstract
    • The beryllium (Be) main chamber wall interaction during a 5 MA/1.8 T upward, unmitigated VDE scenario, first analysed in [J. Coburn et al., Phys. Scr. T171 (2020) 014076] for ITER, has been re-evaluated using the latest energy deposition analysis software. Updates to the DINA disruption model are summarized, including an improved numerical convergence for the OD power balance, limitations on the safety factor within the plasma core, and the choice to maintain a constant plasma + halo poloidal cross-section. Such updates result in a broad halo region and higher radiated power fractions compared to previous models. The new scenario lasts for similar to 75 ms and deposits similar to 29 MJ of energy, with the radial distribution of parallel heat flux q parallel to(r) resembling an exponential falloff with an effective lambda(q) = 75 -198 mm. A maximum halo width w(h) of 0.52 m at the outboard midplane is observed. SMITER field line tracing and energy deposition simulations calculate a q(perpendicular to,max) of similar to 83 MW/m(2) on the upper first wall panels (FWP). Heat transfer calculations with the MEMOS-U code show that the FWP surface temperature reaches similar to 1000 K, well below the Be melt threshold. Variations of this 5 MA scenario with Be impurity densities from 0 to 3.10(19) m(-3) also remain below the melt threshold despite differences in energy deposition and duration. These results are in contrast to the early study which predicted melt damage to the first wall [J. Coburn et al., Phys. Scr. T171 (2020) 014076], and emphasize the importance of accurate models for the halo width w(h) and the heat flux distribution q parallel to(r) within that halo width. The 2020 halo model in DINA has been compared with halo current experiments on COMPASS, JET, and Alcator C-Mod, and the preliminary results build confidence in the broad halo width predictions. Results for the 5 MA VDE are compared with those for a 15 MA equivalent, generated using the new DINA model. At the higher current, significant melting of the upper FWP is to be expected.
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14.
  • Corre, Y., et al. (författare)
  • Testing of ITER-grade plasma facing units in the WEST tokamak: Progress in understanding heat loading and damage mechanisms
  • 2023
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
  • Tidskriftsartikel (refereegranskat)abstract
    • Assessing the performance of the ITER design for the tungsten (W) divertor Plasma Facing Units (PFUs) in a tokamak environment is a high priority issue to ensure efficient plasma operation. This paper reviews the most recent results derived from experiments and post-mortem analysis of the ITER-grade PFUs exposed in the WEST tokamak and the associated modelling, with a focus on understanding heat loading and damage evolution. Several shaping options, sharp or chamfered leading edge (LE), unshaped or shaped blocks with a toroidal bevel as foreseen in ITER, were investigated, under steady state heat fluxes of up to 120 MW⋅m−2 and 6 MW⋅m−2 on the sharp LE and top surface of the block, respectively. A very high spatial resolution (VHR) infrared (IR) camera (0.1 mm/pixel) was used to derive the temporal and surface distribution of the temperature and heat load on the castellated tungsten blocks for different geometric alignment and plasma conditions. Photonic modelling was required to reproduce the IR measurements in particular in the toroidal and poloidal gaps of the mono-block (MB) stacks where high apparent temperatures are observed. Specular reflection is found to be the dominant emitter in these parts of the blocks. W-cracking was observed on the leading edge of the blocks already within the first phase of plasma operation, during which the divertor was equipped with unshaped PFUs, including some intentionally misaligned blocks. Numerical analysis taking into account softening processes and mechanical stresses, revealed brittle failure due to transients as the dominant failure mechanisms. Ductile failure was observed in one particular block used for the melting experiment, therefore under extremely high steady state heat load conditions. W-melting achieved on actively cooled PFU exhibits specific features: shallow melting and slow melt displacement. Plasma exposure of pre-damaged PFUs at various damage levels (crack network and melted droplets) was carried out under high heat load conditions with several hours of cumulated plasma duration. IR data and preliminary surface analyses show no evidence of significant degradation damage progression under these conditions.
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15.
  • Cupak, C., et al. (författare)
  • Absence of synergistic effects in quasi-simultaneous sputtering of tungsten by Ar and D ions
  • 2023
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 35
  • Tidskriftsartikel (refereegranskat)abstract
    • A quartz crystal microbalance was used to experimentally study the erosion of tungsten during rapidly alternating bombardment with 2 keV argon and deuterium projectiles. A key goal was to investigate whether the mean sputtering yield of the alternating irradiation can be predicted from data for sputtering yields of single ion species. In addition, influences by residual gas pressure in the UHV experiment and variable ion fluxes have been studied. Our results show that the mean sputtering yield of irradiations with alternating ion species can be well predicted for a range of different fluence ratios as a simple superposition of individual sputtering yields, weighted by the respective relative fluences. This finding supports that no synergistic sputtering effects were relevant in the investigated low-flux regime.
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16.
  • Cupak, C., et al. (författare)
  • Retention of deuterium in beryllium : A combined investigation using TDS, ERDA and EBS
  • 2022
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 33
  • Tidskriftsartikel (refereegranskat)abstract
    • We have studied the retention of deuterium in beryllium, implanted with an energy of 500 eV/D, using a combination of thermal desorption spectroscopy, elastic recoil detection analysis and elastic backscattering spectroscopy. The parallel use of these techniques allowed us to directly quantify the absolute deuterium content reduction of the sample for specific desorption peaks observed during thermal annealing. In addition, the presence of a beryllium oxide surface layer was observed, despite sputter-cleaning of the sample was initially conducted in-situ. A main result was that similar to 85 % of the retained deuterium got released during the primary desorption peak at 400 K. A smaller, secondary desorption peak was identified at 540 K. All deuterium could be removed from the Be sample by heating it to a temperature of 800 K.
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17.
  • De Angeli, M., et al. (författare)
  • Cross machine investigation of magnetic tokamak dust; structural and magnetic analysis
  • 2021
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 28
  • Tidskriftsartikel (refereegranskat)abstract
    • Magnetic dust collected from multiple fusion devices (FTU, Alcator C-Mod, COMPASS) that feature different plasma-facing components (PFCs) and toroidal magnetic fields has been analyzed by means of the X-ray diffraction technique aiming to investigate the nature and origin of dust magnetism. Analysis led to the conclusion that the main mechanism of ferromagnetic dust formation is the change of iron crystalline phase from austenitic to ferritic during the re-solidification of stainless steel droplets. Analysis also revealed differences in the collected dust structure and an unexpectedly high amount of stainless steel based dust in its native austenitic phase. Theoretical estimates showed that the magnetic moment force can also mobilize strongly paramagnetic adhered dust prior to the establishment of proper tokamak discharges. The post-mortem analysis of dust collected during pure magnetic discharges in FTU confirmed these estimates.
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18.
  • De Angeli, M., et al. (författare)
  • Post-mortem and in-situ investigations of magnetic dust in ASDEX Upgrade
  • 2023
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 36
  • Tidskriftsartikel (refereegranskat)abstract
    • Pre-plasma mobilization of magnetic dust can be an important issue for future fusion reactors where plasma breakdown is critical. A combined on-line and off-line study of magnetic dust in ASDEX Upgrade is reported. Post-mortem collection revealed similar composition and morphology compared to other tokamaks, but the overall amount was much smaller. Optical and IR camera diagnostics excluded dust flybys prior to plasma start-up. The negative detection is discussed in light of the magnetic dust properties, the strength of mobilizing forces and the temporal evolution of the magnetic field.
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19.
  • De Angeli, M., et al. (författare)
  • Remobilization of tungsten dust from castellated plasma-facing components
  • 2017
  • Ingår i: NUCLEAR MATERIALS AND ENERGY. - : Elsevier BV. - 2352-1791. ; 12, s. 536-540
  • Tidskriftsartikel (refereegranskat)abstract
    • Studies of tungsten dust remobilization from castellated plasma-facing components can shed light to whether gaps constitute a dust accumulation site with important implications for monitoring but also removal. Castellated structures of ITER relevant geometry that contained pre-adhered tungsten dust of controlled deposition profile have been exposed in the Pilot-PSI linear device. The experiments were performed under steady state and transient plasma conditions, as well as varying magnetic field topologies. The results suggest that dust remobilization from the plasma-facing monoblock surface can enhance dust trapping in the gaps and that tungsten dust is efficiently trapped inside the gaps.
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20.
  • De Temmerman, Gregory, et al. (författare)
  • Data on erosion and hydrogen fuel retention in Beryllium plasma-facing materials
  • 2021
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 27
  • Tidskriftsartikel (refereegranskat)abstract
    • ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m(2). Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma-wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.
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21.
  • Denis, J., et al. (författare)
  • Dynamic modelling of local fuel inventory and desorption in the whole tokamak vacuum vessel for auto-consistent plasma-wall interaction simulations
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 550-557
  • 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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22.
  • Dittrich, Laura, et al. (författare)
  • Impact of ion irradiation and film deposition on optical and fuel retention properties of Mo polycrystalline and single crystal mirrors
  • 2023
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
  • Tidskriftsartikel (refereegranskat)abstract
    • Polycrystalline (PC) and single crystal (SC) molybdenum mirrors were irradiated with 98Mo+, 1H+, 4He+, 11B+ and 184W+. Energies were chosen to impact the optically active region (up to 30 nm deep) of Mo mirrors. Some surfaces were coated by magnetron sputtering either with B or W films 4–65 nm thick. The overall objective was to simulate the neutron-induced damage and transmutation (H, He), and the impact of H, He, B, W on the optical performance of test mirrors, and on fuel retention. In parallel, a set of PC Mo mirrors irradiated with 1.6 MeV 98Mo3+ to a damage of 2 dpa and 20 dpa was installed in the JET tokamak for exposure during deuterium-tritium campaigns. Data from spectrophotometric, ion beam and microscopy techniques reveal: (i) the irradiation decreased specular reflectivity, whereby the differences between PC and SC in reflectivity are very small, (ii) He is retained in bubbles within 25–30 nm of the subsurface layer in all irradiated materials, (iii) W, either deposited or implanted, decreases reflectivity, but the strongest reflectivity degradation is caused by B deposition. Laboratory studies show the correlation of damage and H retention. Several cycles of W deposition and its removal from SC-Mo mirrors by plasma-assisted methods were also performed.
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23.
  • Eich, T., et al. (författare)
  • ELM divertor peak energy fluence scaling to ITER with data from JET, MAST and ASDEX upgrade
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 84-90
  • Tidskriftsartikel (refereegranskat)abstract
    • A newly established scaling of the ELM energy fluence using dedicated data sets from JET operation with CFC & ILW plasma facing components (PFCs), ASDEX Upgrade (AUG) operation with both CFC and full-W PFCs and MAST with CFC walls has been generated. The scaling reveals an approximately linear dependence of the peak ELM energy with the pedestal top electron pressure and with the minor radius; a square root dependence is seen on the relative ELM loss energy. The result of this scaling gives a range in parallel peak ELM energy fluence of 10-30 MJm(-2) for ITER Q = 10 operation and 2.5-7.5 MJm(-2) for intermediate ITER operation at 7.5 MA and 2.65 T. These latter numbers are calculated using a numerical regression (epsilon(II) = 0.28 MJ/m(2) n(e)(0.75) T-e(1) Delta E-ELM(0.5) R-1(geo)). A simple model for ELM induced thermal load is introduced, resulting in an expression for the ELM energy fluence of epsilon(II) congruent to 6 pi p(e) R-geo q(edge). The relative ELM loss energy in the data is between 2-10% and the ELM energy fluence varies within a range of 10(0.5) similar to 3 consistently for each individual device. The so far analysed power load database for ELM mitigation experiments from JET-EFCC and Kicks, MAST-RMP and AUG-RMP operation are found to be consistent with both the scaling and the introduced model, ie not showing a further reduction with respect to their pedestal pressure. The extrapolated ELM energy fluencies are compared to material limits in ITER and found to be of concern.
  •  
24.
  • Fellinger, Joris, et al. (författare)
  • Tungsten based divertor development for Wendelstein 7-X
  • 2023
  • Ingår i: Nuclear Materials and Energy. - 2352-1791. ; 37
  • Tidskriftsartikel (refereegranskat)abstract
    • Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified. This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx. < 0.25 m2 and < 60 kg) using the existing water cooling infrastructure providing 5 l/s and roughly maximum 15 bar pressure drop per module. The main technology under qualification is based on a CuCrZr heat sink made either by additive manufacturing using laser powder bed fusion (LPBF) or by uniaxial diffusion welding of pre-machined forged CuCrZr plates. After heat treatment, the plasma-facing side of the heat sink is covered by W or if feasible by the more ductile WNiFe, preferably by coating or alternatively by hot isostatic pressing W based tiles with a soft OFE-Cu interlayer. Last step is a final machining of the plasma-exposed surface and the interfaces to the water supply lines and supports to correct manufacturing deformations.
  •  
25.
  • Fortuna-Zalesna, E., et al. (författare)
  • Studies of dust from JET with the ITER-Like Wall : Composition and internal structure
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 12, s. 582-587
  • Tidskriftsartikel (refereegranskat)abstract
    • Results are presented for the dust survey performed at JET after the second experimental campaign with the ITER-Like Wall: 2013-2014. Samples were collected on adhesive stickers from several different positions in the divertor both on the tiles and on the divertor carrier. Brittle dust-forming deposits on test mirrors from the inner divertor wall were also studied. Comprehensive characterization accomplished by a wide range of high-resolution microscopy techniques, including focused ion beam, has led to the identification of several classes of particles: (i) beryllium flakes originating either from the Be coatings from the inner wall cladding or Be-rich mixed co-deposits resulting from material migration; (ii) beryllium droplets and splashes; (iii) tungsten and nickel-rich (from Inconel) droplets; (iv) mixed material layers with a various content of small (8-200 nm) W-Mo and Ni-based debris. A significant content of nitrogen from plasma edge cooling has been identified in all types of co-deposits. A comparison between particles collected after the first and second experimental campaign is also presented and discussed. (C) 2016 Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license.
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26.
  • Garcia Carrasco, Alvaro, 1989-, et al. (författare)
  • Investigation of probe surfaces after ion cyclotron wall conditioning in ASDEX upgrade
  • 2017
  • Ingår i: NUCLEAR MATERIALS AND ENERGY. - : Elsevier. - 2352-1791. ; 12, s. 733-735
  • Tidskriftsartikel (refereegranskat)abstract
    • For the first time, material analysis techniques have been applied to study the effect of ion cyclotron wall conditioning (ICWC) on probe surfaces in a metal-wall machine. ICWC is a technique envisaged to contribute to the removal of fuel and impurities from the first wall of ITER. The objective of this work was to assess impurity migration under ICWC operation. Tungsten probes were exposed in ASDEX Upgrade to discharges in helium. After wall conditioning, the probes were covered with a co-deposited layer containing D, B, C, N, O and relatively high amount of He. The concentration ratio He/C+B was 0.7. The formation of the co-deposited layer indicates that a fraction of the impurities desorbed from the wall under ICWC operation are transported by plasma and deposited away from their original location.
  •  
27.
  • Garcia Carrasco, Alvaro, 1989-, et al. (författare)
  • Plasma impact on diagnostic mirrors in JET
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 12, s. 506-512
  • Tidskriftsartikel (refereegranskat)abstract
    • Metallic mirrors will be essential components of all optical systems for plasma diagnosis in ITER. This contribution provides a comprehensive account on plasma impact on diagnostic mirrors in JET with the ITER-Like Wall. Specimens from the First Mirror Test and the lithium-beam diagnostic have been studied by spectrophotometry, ion beam analysis and electron microscopy. Test mirrors made of molybdenum were retrieved from the main chamber and the divertor after exposure to the 2013-2014 experimental campaign. In the main chamber, only mirrors located at the entrance of the carrier lost reflectivity (Be deposition), while those located deeper in the carrier were only slightly affected. The performance of mirrors in the JET divertor was strongly degraded by deposition of beryllium, tungsten and other species. Mirrors from the lithium-beam diagnostic have been studied for the first time. Gold coatings were severely damaged by intense arcing. As a consequence, material mixing of the gold layer with the stainless steel substrate occurred. Total reflectivity dropped from over 90% to less than 60%, i.e. to the level typical for stainless steel. (C) 2017 Elsevier Ltd. This is an open access article under the CC BY-NC-ND license.
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28.
  • Gaspar, J., et al. (författare)
  • Heat flux analysis of Type-I ELM impact on a sloped, protruding surface in the JET bulk tungsten divertor
  • 2018
  • Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 17, s. 182-187
  • Tidskriftsartikel (refereegranskat)abstract
    • Tungsten (W) melting due to transient power loads, for example those delivered by edge localised modes (ELMs), is a major concern for next step fusion devices. A series of experiments has been performed on JET to investigate the dynamics of Type-I ELM-induced transient melting. Following initial exposures in 2013 of a W-lamella with sharp leading edge in the bulk W outer divertor, new experiments have been performed in 2016-2017 on a protruding W-lamella with a 15 degrees slope, allowing direct and spatially resolved (0.85 mm/pixel) observation of the top surface using the IR thermography system viewing from the top of the poloidal cross-section. Thermal and IR analysis have already been conducted assuming the geometrical projection of the parallel heat flux on the W-lamellas, thus ignoring the gyro-radius orbit of plasma particles. Although it is well justified during L-mode or inter-ELM period, the hypothesis becomes questionable during ELM when the ion Larmor radius is larger. The goal of this paper is to extend the previous analysis based on the forward approach to the H-mode discharges and investigate in particular the gyro-radius effect during the Type-I ELMs, those used to achieve transient melting on the slope of the protruding W-lamella. Surface temperatures measured by the IR camera are compared with reconstructed synthetic data from 3D thermal modelling using heat loads derived from optical projection of the parallel heat flux (ignoring the gyro-radius orbit), 2D gyro-radius orbit and particle-in-cell (PIC) simulations describing the influence of finite Larmor-radius effects and electrical potential on the deposited power flux. Results show that the ELM power deposition behaves differently than the optical projection of the parallel heat flux, contrary to the L-mode observations, and may thus be due to the much larger gyro-orbits of the energetic ELM ions in comparison to L-mode or inter-ELM conditions.
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29.
  • Guillemaut, C., et al. (författare)
  • Main chamber wall plasma loads in JET-ITER-like wall at high radiated fraction
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 234-240
  • Tidskriftsartikel (refereegranskat)abstract
    • Future tokamak reactors of conventional design will require high levels of exhaust power dissipation (more than 90% of the input power) if power densities at the divertor targets are to remain compatible with active cooling. Impurity seeded H-mode discharges in JET-ITER-like Wall (ILW) have reached a maximum radiative fraction (F-rad) of similar to 75%. Divertor Langmuir probe (LP) measurements in these discharges indicate, however, that less than similar to 3% of the thermal plasma power reaches the targets, suggesting a missing channel for power loss. This paper presents experimental evidence from limiter LP for enhanced cross-field particle fluxes on the main chamber walls at high F-rad. In H-mode nitrogen-seeded discharges with F-rad increasing from similar to 30% to up to similar to 75%, the main chamber wall particle fluence rises by a factor similar to 3 while the divertor plasma fluence drops by one order of magnitude. Contribution of main chamber wall particle losses to detachment, as suggested by EDGE2D-EIRENE modeling, is not sufficient to explain the magnitude of the observed divertor fluence reduction. An intermediate detached case obtained at F-rad similar to 60% with neon seeding is also presented. Heat loads were measured using the main chamber wall thermocouples. Comparison between thermocouple and bolometry measurements shows that the fraction of the input power transported to the main chamber wall remains below similar to 5%, whatever the divertor detachment state is. Main chamber sputtering of beryllium by deuterium is reduced in detached conditions only on the low field side. If the fraction of power exhaust dissipated to the main chamber wall by cross-field transport in future reactors is similar to the JET-ILW levels, wall plasma power loading should not be an issue. However, other contributions such as charge exchange may be a problem.
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30.
  • Hansson, Niklas, 1992, et al. (författare)
  • Alpha dose rate calculations for UO2 based materials using stopping power models
  • 2020
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 22
  • Tidskriftsartikel (refereegranskat)abstract
    • Accurate dose rate models for UO2 based materials in contact with water are important in the modeling of the radiolytically promoted dissolution of spent fuel. Dose rates of α-doped UO2 and un-irradiated MOX fuel were modelled using the ASTAR and SRIM stopping power databases. Dose rates were calculated as a function of distance from the active surface. Comparisons with common dose rate calculation models and the combined Bethe-Bloch and Lindhard–Scharff (LS) equation were performed. It was shown that the ASTAR and SRIM databases could more accurately simulate an α-spectrum compared to the Bethe-Bloch-LS equation. A comparison between the continuous slowing down approximation (CSDA) and the radial projection algorithm in the SRIM program was performed, and it was shown that CSDA overestimates the range of the α-particles by a few percent. This leads to an overestimation of the α-dose rate at distances close to the maximum range of the α-particle in water. A relationship between the average dose rate to specific α-activity ratio as a function of α-energy was obtained from the calculations, which can easily be implemented in alpha dose rate calculations of a UO2 based materials.
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31.
  • Hatano, Y., et al. (författare)
  • Tritium distributions on W-coated divertor tiles used in the third JET ITER-like wall campaign
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 258-261
  • 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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32.
  • Heinola, K., et al. (författare)
  • Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 19, s. 397-402
  • 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.
  •  
33.
  • Huber, A., et al. (författare)
  • Comparative H-mode density limit studies in JET and AUG
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 100-110
  • Forskningsöversikt (refereegranskat)abstract
    • Identification of the mechanisms for the H-mode density limit in machines with fully metallic walls, and their scaling to future devices is essential to find for these machines the optimal operational boundaries with the highest attainable density and confinement. Systematic investigations of H-mode density limit plasmas in experiments with deuterium external gas fuelling have been performed on machines with fully metallic walls, JET and AUG and results have been compared with one another. Basically, the operation phases are identical for both tokamaks: the stable H-mode phase, degrading H-mode phase, breakdown of the H-mode with energy confinement deterioration usually accompanied by a dithering cycling phase, followed by the l -mode phase. The observed H-mode density limit on both machines is found close to the Greenwald limit (n/n GW = 0.8-1.1 in the observed magnetic configurations). The similar behavior of the radiation on both tokamaks demonstrates that the density limit (DL) is neither related to additional energy losses from the confined region by radiation, nor to an inward collapse of the hot discharge core induced by overcooling of the plasma periphery by radiation. It was observed on both machines that detachment, as well as the X-point MARFE itself, does not trigger a transition in the confinement regime and thus does not present a limit on the plasma density. It is the plasma confinement, most likely determined by edge parameters, which is ultimately responsible for the transition from H-to L-mode. The measured Greenwald fractions are found to be consistent with the predictions from different theoretical models [16,30] based on MHD instability theory in the near-SOL.
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34.
  • Huber, A., et al. (författare)
  • Determination of tungsten sources in the JET-ILW divertor by spectroscopic imaging in the presence of a strong plasma continuum
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 18, s. 118-124
  • 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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35.
  • Journeau, Christophe, et al. (författare)
  • Transient interactions of boron carbide with molten uranium oxide
  • 2021
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 29
  • Tidskriftsartikel (refereegranskat)abstract
    • Thermodynamic equilibrium calculations have shown that, at the elevated temperatures anticipated during se-vere accidents, uranium oxide-boron carbide interactions are likely to result in significant boron volatilization and the formation of uranium borides. This interaction is further investigated through transient experiments whereby boron carbide pellets are submerged in uranium dioxide melts held at elevated temperatures beneath inert atmospheres. X-ray diffraction and electron-microscopy analyses revealed the aerosols generated by the interaction to be rich in boron sesquioxide (B2O3) and uranium oxide. The solidified melt ingots recovered post-test exhibited highly heterogeneous chemistry with some samples richer in UO2 and others richer in U3O8. Thermodynamic equilibrium calculations indicate that this could be explained by heterogeneity in redox envi-ronment. The second, larger-scale, interaction, with an increased boron inventory, indicated uranium oxide -boron carbide reaction formed either uranium boride or uranium borocarbide.
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36.
  • Kirschner, Andreas, et al. (författare)
  • Modelling of deposition and erosion of injected WF6 and MoF6 in TEXTOR
  • 2016
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791.
  • Tidskriftsartikel (refereegranskat)abstract
    • Tracer injection experiments in TEXTOR with MoF 6 and WF 6 lead to local deposition of about 6% for Mo and about 1% for W relative to the injected amount of Mo and W atoms. Modelling of these experiments has been done with ERO applying updated data for physical sputtering. The dissociation of the injected molecules has been treated in a simplified manner due to the lack of dissociation rate coefficients. How- ever, with this it was possible to reproduce the observed radial penetration of Mo and W atoms into the plasma. The modelled local deposition efficiencies are about 50% for Mo and 60% for W assuming typical plasma parameters for the experimental conditions used. To reproduce the measured deposition efficien- cies an enhancement factor for the erosion of deposited Mo and W has to be assumed ( ∼10 for Mo and ∼25 for W). Due to the rather low electron temperature T e of these plasma conditions (T e ∼15 eV at the location of injection), Mo and W are mostly sputtered by impurities whereas sputtering due to deuterium is negligible. A parameter study applying larger electron temperature leads to increased sputtering and thus to reduced local deposition efficiencies of about 30% for Mo and 5% for W. Though, even under these conditions enhanced erosion, albeit with reduced enhancement factors, is needed in the modelling to obtain the small measured deposition efficiencies.
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37.
  • Kirschner, A., et al. (författare)
  • Modelling of tungsten erosion and deposition in the divertor of JET-ILW in comparison to experimental findings
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 18, s. 239-244
  • 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.
  •  
38.
  • Kit, A., et al. (författare)
  • Developing deep learning algorithms for inferring upstream separatrix density at JET
  • 2023
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 34
  • Tidskriftsartikel (refereegranskat)abstract
    • Predictive and real-time inference capability for the upstream separatrix electron density, ne, sep, is essential for design and control of core-edge integrated plasma scenarios. In this study, both supervised and semi -supervised machine learning algorithms are explored to establish direct mapping as well as indirect compressed representation of the pedestal profiles for predictions and inference of ne, sep. Based on the EUROfusion pedestal database for JET (Frassinetti et al., 2021), a tabular dataset was created, consisting of machine parameters, fraction of ELM cycle, high resolution Thomson scattering profiles of electron density and temperature, and ne, sep for 608 JET shots. Using the tabular dataset, the direct mapping approach provides a mapping of machine parameters and ELM percentage to ne, sep. Through representation learning, a compressed representation of the experimental pedestal electron density and temperature profiles is established. By conditioning the representation with machine control parameters, a probabilistic generative predictive model is established. For prediction, the machine parameters can be used to establish a conditional distribution of the compressed pedestal profiles, and the decoder that is trained as part of the algorithm can be used to decode the compressed representation back to full pedestal profiles. Although, in this work, a proof-of-principle for predicting and inferring ne, sep is given, such a representation learning can be used also for many other applications as the full pedestal profile is predicted. An implementation of this work can be found at https://github.com/ fusionby2030/psi_2022.
  •  
39.
  • Kolbas, Daria, et al. (författare)
  • Role of various influencing parameters on high temperature fretting behaviour of different tribopairs in liquid lead
  • 2024
  • Ingår i: Nuclear Materials and Energy. - : Elsevier Ltd. - 2352-1791. ; 40
  • Tidskriftsartikel (refereegranskat)abstract
    • The increasing interest in liquid metal cooled nuclear reactors provides technical and scientific challenges such as the understanding, prevention, and prediction of the degradation of materials in liquid lead. Critical components include the fuel rods, heat exchanger tubes, and pump impellers. These functional elements are exposed to mechanical loading (up to 40 MPa), high temperatures (450–550 °C), and fluid-induced vibrations (up to 25 Hz). Under such conditions, fretting wear occurs between e.g., the spacer wire and the outer surface of the fuel or heat exchanger tubes. This work is aimed to establish a laboratory-scale fretting wear test setup and develop test methodology to enable systematic material characterisation in liquid metal environments. The results obtained by using the described methodology indicate that adhesive wear is the dominant degradation mechanism, and 316L stainless steel shows a higher coefficient of friction but a lower wear volume/tribolayer volume compared to 100Cr6 bearing steel. These results are in agreement with those reported in open literature and demonstrates the suitability of the presented method for conducting fretting tests and analysis for various materials and contact configurations in liquid lead environment.
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40.
  •  
41.
  • Krat, S., et al. (författare)
  • Beryllium film deposition in cavity samples in remote areas of the JET divertor during the 2011-2012 ITER-like wall campaign
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 548-552
  • Tidskriftsartikel (refereegranskat)abstract
    • Beryllium film deposition was studied with cavity samples in remote areas of the inner and outer JET divertor and below divertor tile 5 during the 2011-2012 campaign with the ITER-like wall. Predominantly beryllium films were formed inside the cavities with some additional carbon, the ratio Be/C was > 2. These deposited layers had high D/(Be+C) ratios of about 0.3. The formation of these films is mainly due to sticking of beryllium-containing particles with low sticking coefficients < 0.5. The observed surface loss probabilities depend on the position in the divertor. The particles responsible for film deposition originated from the location of in the divertor strike points.
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42.
  • Krat, S., et al. (författare)
  • Erosion at the inner wall of JET during the discharge campaign 2013-2014
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 11, s. 20-24
  • Tidskriftsartikel (refereegranskat)abstract
    • The erosion of Be and W marker layers was investigated using long-term samples containing marker layers during the second ITER-like wall discharge campaign 2013-2014 (ILW-2). The samples were mounted in Be coated Inconel tiles between the inner wall guard limiters (IWGL). They were analyzed using elastic backscattering (EBS) before and after exposure. All samples showed noticeable erosion. The results were compared to the data for Be and W erosion rates for the first 2011-2012 JET ITER-like wall (ILW-1) campaign, and to the data for C erosion during the 2005-2009 campaign when JET was operated with a carbon wall. The mean W erosion rates and the toroidal and poloidal distributions of the W erosion were nearly the same for the ILW-1 and ILW-2 campaigns. The mean erosion rate of Be during the ILW-2 campaign was smaller by a factor of about two compared to the ILW-1 campaign. 
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43.
  • Krivska, A., et al. (författare)
  • RF sheath modeling of experimentally observed plasma surface interactions with the JET ITER-Like Antenna
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 19, s. 324-329
  • 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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44.
  • Kumar, M., et al. (författare)
  • Identification of BeO and BeOxDy in melted zones of the JET Be limiter tiles : Raman study using comparison with laboratory samples
  • 2018
  • Ingår i: Nuclear Materials and Energy. - : ELSEVIER. - 2352-1791. ; 17, s. 295-301
  • Tidskriftsartikel (refereegranskat)abstract
    • Beryllium oxide (BeO) and deuteroxide (BeOxDy) have been found on the melted zone of a beryllium tile extracted from the upper dump plate of JET-ILW (2011-2012 campaign). Results have been obtained using Raman microscopy, which is sensitive to both the chemical bond and crystal structure, with a micrometric lateral resolution. BeO is found with a wurtzite crystal structure. BeOxDy is found as three different types which are not the beta-phase but behaves as molecular species like Be(OD)(2), O(Be-D)(2) and DBeOD. The presence of a small amount of trapped D2O is also suspected. Our results therefore strongly suggest that D trapping occurs after melting through the formation of deuteroxides. The temperature increase favors the formation of crystal BeO which favors deuterium trapping through OD bonding.
  •  
45.
  • Kumpulainen, H. A., et al. (författare)
  • ELM and inter-ELM tungsten erosion sources in high-power, JET ITER-like wall H-mode plasmas
  • 2022
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 33
  • Tidskriftsartikel (refereegranskat)abstract
    • Simulations of JET ITER-like wall high-confinement mode plasmas, including type-I edge-localised modes (ELMs), using JINTRAC for the background plasmas and ERO2.0 for tungsten erosion and transport, predict virtually perfect screening of the primary W erosion sources at the divertor targets during both the ELM and inter-ELM phases. The largest source of W influx to the main plasma is predicted to be the outer vertical divertor due to sputtering by energetic fuel (D, T) atoms from charge-exchange reactions. ERO2.0 predictions accurately reproduce the measured W I emission in the low-field side divertor, but underpredict the W II emission by a factor of 10. Potential reasons for the W II discrepancy include uncertainties in the atomic data, assumptions on the sheath properties and the sputtering angle distribution, and the impact of metastable states.
  •  
46.
  • Lahtinen, A., et al. (författare)
  • Deuterium retention in the divertor tiles of JET ITER-Like wall
  • 2017
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 12, s. 655-661
  • Tidskriftsartikel (refereegranskat)abstract
    • Divertor tiles removed after the second JET ITER-Like Wall campaign 2013-2014 (ILW-2) were studied using Secondary Ion Mass Spectrometry (SIMS). Measurements show that the thickest beryllium (Be) dominated deposition layers are located at the upper part of the inner divertor and are up to similar to 40 mu m thick at the lower part of Tile 0 exposed in 2011-2014. The highest deuterium (D) amounts (>8 . 10 18 at./cm(2)), in contrast, were found on the upper part of Tile 1 (2013-2014), where the Be deposits are similar to 10 mu m thick. D was mainly retained in the near-surface layer of the Be deposits but also deeper in tungsten (W) and molybdenum (Mo) layers of the marker coated tiles, especially at W-Mo layer interfaces. D retention for the ILW-2 divertor tiles is higher than for the first campaign 2011-2012 (ILW-1) and probable reasons for the difference are that SIMS measurements for the ILW-2 samples were done deeper than for the ILW-1 samples, some of the tiles were exposed during both ILW-1 and ILW-2 and therefore had a longer exposure time, and the differences between ILW-1 and ILW-2 campaigns e.g. in strike point distributions and injected powers.
  •  
47.
  • Lee, S. E., et al. (författare)
  • Global distribution of tritium in JET with the ITER-like wall
  • 2021
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 26
  • Tidskriftsartikel (refereegranskat)abstract
    • Nondestructive analysis of tritium (T) distribution was performed by means of imaging plate technique on specimens cut from the Be limiters, W-coated carbon tiles and bulk W lamellae retrieved from the JET tokamak after the first and third experimental campaigns with the ITER-like wall. Afterwards, analyses were continued using X-ray photoelectron spectroscopy, microscopy techniques and thermal desorption spectroscopy. Co-deposits formed on the W-coated tiles in the 1st campaign showed large T retention because of high carbon content reaching up to 50 atomic %, while the carbon fraction in co-deposits after the 3rd campaign was distinctly lower. The T retention of the plasma-facing surface of the bulk W tile was smaller than that of the W-coated tiles by a factor of 20, while deposition of small amount of T was found at the side surfaces facing to the gaps in a lamella structure. The correlation of T distributions with surface morphology and the discharge conditions is discussed.
  •  
48.
  • Likonen, J., et al. (författare)
  • Investigation of deuterium trapping and release in the JET divertor during the third ILW campaign using TDS
  • 2019
  • Ingår i: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 19, s. 300-306
  • 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.
  •  
49.
  • Likonen, J., et al. (författare)
  • 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
  • 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.
  •  
50.
  • Lindgren, Kristina, 1989, et al. (författare)
  • Elemental distribution in a decommissioned high Ni and Mn reactor pressure vessel weld metal from a boiling water reactor
  • 2023
  • Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; , s. 101466-101466
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper, weld metal from unique material of a decommissioned boiling water reactor pressure vessel is investigated. The reactor was in operation for 23 effective full power years. The elemental distribution of Ni, Mn, Si and Cu in the material is analysed using atom probe tomography. There are no well-defined clusters of these elements in the weld metal. However, some clustering tendencies of Ni was found, and these are interpreted as a high number density of small features. Cu atoms were found to statistically be closer to Ni atoms than in a fully random solid solution. The impact of the non-random elemental distribution on mechanical properties is judged to be limited.
  •  
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