| 1. |
- Labit, B., et al.
(författare)
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Dependence on plasma shape and plasma fueling for small edge-localized mode regimes in TCV and ASDEX Upgrade
- 2019
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:8
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Tidskriftsartikel (refereegranskat)abstract
- © 2019 Institute of Physics Publishing. All rights reserved. Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (ne,sep/nG ∼ 0.3), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened.
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| 2. |
- Bombarda, F., et al.
(författare)
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Runaway electron beam control
- 2019
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Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 61:1
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Tidskriftsartikel (refereegranskat)
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| 3. |
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- 2018
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:1
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Forskningsöversikt (refereegranskat)
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| 4. |
- Allen, S. L., et al.
(författare)
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C-13 transport studies in L-mode divertor plasmas on DIII-D
- 2005
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 337-39:03-jan, s. 30-34
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Tidskriftsartikel (refereegranskat)abstract
- (CH4)-C-13 was injected with a toroidally-symmetric gas system into 22 identical lower-single-null L-mode discharges on DIII-D. The injection level was adjusted so that it did not significantly perturb the core or divertor plasmas, with a duration of similar to 3 s on each shot, for a total of similar to 300 T L of injected particles. The plasma shape remained very constant; the divertor strike points were controlled to similar to 1 cm at the divertor plate. At the beginning of the subsequent machine vent, 29 carbon tiles were removed for nuclear reaction analysis of C-13 content to determine regions of carbon deposition. It was found that only the tiles inboard of the inner strike point had appreciable 1 3 C above background. Visible spectroscopy measurements of the carbon injection and comparisons with modeling are consistent with carbon transport by means of scrape-off layer flow.
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| 5. |
- Andrew, P, et al.
(författare)
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Outer divertor target deposited layers during reversed magnetic field operation in JET
- 2005
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 337:1-3, s. 99-103
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Tidskriftsartikel (refereegranskat)abstract
- Divertor surface temperatures are significantly affected by the presence of deposited surface layers. This phenomenon can be used to monitor deposited layer evolution on a shot-by-shot basis. It was found that during an experimental campaign where the B x del B direction was reversed that the outer target, normally an erosion zone, became a deposition zone.
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| 6. |
- Arnoux, G., et al.
(författare)
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Thermal analysis of an exposed tungsten edge in the JET divertor
- 2015
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 463, s. 415-419
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Tidskriftsartikel (refereegranskat)abstract
- In the recent melt experiments with the JET tungsten divertor, we observe that the heat flux impacting on a leading edge is 3-10 times lower than a geometrical projection would predict. The surface temperature, tungsten vaporisation rate and melt motion measured during these experiments is consistent with the simulations using the MEMOS code, only if one applies the heat flux reduction. This unexpected observation is the result of our efforts to demonstrate that the tungsten lamella was melted by ELM induced transient heat loads only. This paper describes in details the measurements and data analysis method that led us to this strong conclusion. The reason for the reduced heat flux are yet to be clearly established and we provide some ideas to explore. Explaining the physics of this heat flux reduction would allow to understand whether it can be extrapolated to ITER.
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| 7. |
- Baron-Wiechec, A., et al.
(författare)
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First dust study in JET with the ITER-like wall : sampling, analysis and classification
- 2015
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:11
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Tidskriftsartikel (refereegranskat)abstract
- Results of the first dust survey in JET with the ITER-Like Wall (JET-ILW) are presented. The sampling was performed using adhesive stickers from the divertor tiles where the greatest material deposition was detected after the first JET-ILW campaign in 2011-2012. The emphasis was especially on sampling and analysis of metal particles (Be and W) with the aim to determine the composition, size, surface topography and internal dust structure using a large set of methods: high-resolution scanning and transmission electron microscopy, focused ion beam, electron diffraction and also wavelength and energy dispersive x-ray spectroscopy. The most important was the identification of beryllium dust both in the form of flakes and droplets with dimensions in the micrometer range. Tungsten, molybdenum, inconel constituents were identified along with many impurity species. The particles are categorised and the origin of the various constituents discussed.
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| 8. |
- Baron-Wiechec, A., et al.
(författare)
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Global erosion and deposition patterns in JET with the ITER-like wall
- 2015
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 463, s. 157-161
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Tidskriftsartikel (refereegranskat)abstract
- A set of Be and W tiles removed after the first ITER-like wall campaigns (JET-ILW) from 2011 to 2012 has been analysed. The results indicate that the primary erosion site is in the main chamber (Be) as in previous carbon campaigns (JET-C). In particular the limiters tiles near the mid-plane are eroded probably during the limiter phases of discharges. W is found at low concentrations on all plasma-facing surfaces of the vessel indicating deposition via plasma transport initially from the W divertor and from main chamber W-coated tiles; there are also traces of Mo (used as an interlayer for these coatings). Deposited films in the inner divertor have a layered structure, and every layer is dominated by Be with some W and O content.
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| 9. |
- Beringer, J., et al.
(författare)
-
REVIEW OF PARTICLE PHYSICS Particle Data Group
- 2012
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Ingår i: Physical Review D. - 1550-7998 .- 1550-2368. ; 86:1, s. 010001-
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Forskningsöversikt (refereegranskat)abstract
- This biennial Review summarizes much of particle physics. Using data from previous editions, plus 2658 new measurements from 644 papers, we list, evaluate, and average measured properties of gauge bosons, leptons, quarks, mesons, and baryons. We summarize searches for hypothetical particles such as Higgs bosons, heavy neutrinos, and supersymmetric particles. All the particle properties and search limits are listed in Summary Tables. We also give numerous tables, figures, formulae, and reviews of topics such as the Standard Model, particle detectors, probability, and statistics. Among the 112 reviews are many that are new or heavily revised including those on Heavy-Quark and Soft-Collinear Effective Theory, Neutrino Cross Section Measurements, Monte Carlo Event Generators, Lattice QCD, Heavy Quarkonium Spectroscopy, Top Quark, Dark Matter, V-cb & V-ub, Quantum Chromodynamics, High-Energy Collider Parameters, Astrophysical Constants, Cosmological Parameters, and Dark Matter. A booklet is available containing the Summary Tables and abbreviated versions of some of the other sections of this full Review. All tables, listings, and reviews (and errata) are also available on the Particle Data Group website: http://pdg.lbl.gov.
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| 10. |
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| 11. |
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| 12. |
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| 13. |
- Brezinsek, S., et al.
(författare)
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Beryllium migration in JET ITER-like wall plasmas
- 2015
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:6
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Tidskriftsartikel (refereegranskat)abstract
- JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (E-in = 35 eV) and more than 100%, caused by Be self-sputtering (E-in = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at E-in = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.
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| 14. |
- Brezinsek, S., et al.
(författare)
-
Erosion, screening, and migration of tungsten in the JET divertor
- 2019
-
Ingår i: Nuclear Fusion. - : IOP PUBLISHING LTD. - 0029-5515 .- 1741-4326. ; 59:9
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Tidskriftsartikel (refereegranskat)abstract
- The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influx of W into the confined region. The screening of W by the divertor and the transport of W in the plasma determines largely the W content in the plasma core, but the W source strength itself has a vital impact on this process. The JET tokamak experiment provides access to a large set of W erosion-determining parameters and permits a detailed description of the W source in the divertor closest to the ITER one: (i) effective sputtering yields and fluxes as function of impact energy of intrinsic (Be, C) and extrinsic (Ne, N) impurities as well as hydrogenic isotopes (H, D) are determined and predictions for the tritium (T) isotope are made. This includes the quantification of intra- and inter-edge localised mode (ELM) contributions to the total W source in H-mode plasmas which vary owing to the complex flux compositions and energy distributions in the corresponding phases. The sputtering threshold behaviour and the spectroscopic composition analysis provides an insight in the dominating species and plasma phases causing W erosion. (ii) The interplay between the net and gross W erosion source is discussed considering (prompt) re-deposition, thus, the immediate return of W ions back to the surface due to their large Larmor radius, and surface roughness, thus, the difference between smooth bulk-W and rough W-coating components used in the JET divertor. Both effects impact on the balance equation of local W erosion and deposition. (iii) Post-mortem analysis reveals the net erosion/deposition pattern and the W migration paths over long periods of plasma operation identifying the net W transport to remote areas. This W transport is related to the divertor plasma regime, e.g. attached operation with high impact energies of impinging particles or detached operation, as well as to the applied magnetic configuration in the divertor, e.g. close divertor with good geometrical screening of W or open divertor configuration with poor screening. JET equipped with the ITER-like wall (ILW) provided unique access to the net W erosion rate within a series of 151 subsequent H-mode discharges (magnetic field: B-t = 2.0 T, plasma current: I-p = 2.0 MA, auxiliary power P-aux = 12 MW) in one magnetic configuration accumulating 900 s of plasma with particle fluences in the range of 5-6 x 10(26) D(+ )m(-2) in the semi-detached inner and attached outer divertor leg. The comparison of W spectroscopy in the intra-ELM and inter-ELM phases with post-mortem analysis of W marker tiles provides a set of gross and net W erosion values at the outer target plate. ERO code simulations are applied to both divertor leg conditions and reproduce the erosion/deposition pattern as well as confirm the high experimentally observed prompt W re-deposition factors of more than 95% in the intra- and inter-ELM phase of the unseeded deuterium H-mode plasma. Conclusions to the expected divertor conditions in ITER as well as to the JET operation in the DT plasma mixture are drawn on basis of this unique benchmark experiment.
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| 15. |
- Brezinsek, S., et al.
(författare)
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Overview of experimental preparation for the ITER-Like Wall at JET
- 2011
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 415:1, s. S936-S942
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Tidskriftsartikel (refereegranskat)abstract
- Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N-2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 x 10(21) D s(-1) were obtained as references in accompanied gas balance studies.
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| 16. |
- Brezinsek, S., et al.
(författare)
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Residual carbon content in the initial ITER-Like Wall experiments at JET
- 2013
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 438:Suppl., s. S303-S308
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Tidskriftsartikel (refereegranskat)abstract
- The residual carbon content and carbon edge flux in JET have been assessed by three independent diagnostic techniques after start of plasma operation with the ITER-Like Wall (ILW) with beryllium first wall and tungsten divertor: (i) in-situ measurements with optical spectroscopy on low ionisation stages of carbon, (ii) charge-exchange recombination spectroscopy, and (iii) residual gas composition analysis in dedicated global gas balance experiments. Direct comparison experiments in L-mode discharges were carried out between references from the previously installed material configuration with plasma-facing components made of carbon-fibre composite (JET-CFC) and the JET-ILW. The temporal evolution of the C divertor flux since installation of the ILW has been studied in the ohmic phase of dedicated monitoring discharges which have been executed regularly throughout the experimental exploitation so far (60000 plasma seconds). The C flux behaviour in the divertor can be divided in three phases: initial fast drop, moderate reduction phase, and a long lasting phase with almost constant C flux. The Be flux in both divertor legs mirrors the behaviour of C. All experiments and diagnostic techniques demonstrate a strong reduction in C fluxes and C content of more than one order of magnitude with respect to JET-CFC which is in line with the reduction in long-term fuel retention due to co-deposition. There is no evidence of an increase in residual carbon in time, thus no indication that a damage of the thin tungsten coatings on CFC substrate in the divertor occurred.
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| 17. |
- Coad, J. P., et al.
(författare)
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Erosion/deposition in JET during the period 1999-2001
- 2003
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Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 313, s. 419-423
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Tidskriftsartikel (refereegranskat)abstract
- Coated divertor and wall tiles exposed in JET for the 1999-2001 operations have been used to assess erosion/deposition. Deposited films of up to 90 mum thickness at the inner wall of the divertor tiles are, for the most part, enriched in beryllium and other metals, whilst carbon is probably chemically sputtered from these tiles and transported to shadowed regions of the inner divertor. However, from the composition at the surface of the tiles, it appears that the chemical erosion was 'switched off' by reducing the JET vessel wall temperature for the last part of the operations to 200 degreesC. Thick powdery deposits localised at the ion transport limit at each corner of the divertor may be due to physical sputtering. Erosion of the coatings is seen at the outer divertor wall, and on all the inner wall and outer limiter tiles.
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| 18. |
- Coad, J. P., et al.
(författare)
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Overview of material re-deposition and fuel retention studies at JET with the Gas Box divertor
- 2006
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 46:2, s. 350-366
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Tidskriftsartikel (refereegranskat)abstract
- in the period 1998-2001 the JET tokamak was operated with the MkII Gas Box divertor. On two occasions during that period a number of limiter and divertor tiles were retrieved from the torus and then examined ex situ with surface sensitive techniques. Erosion and deposition patterns were determined in order to assess the material erosion, material migration and fuel inventory on plasma facing components. Tracer techniques, e.g. injection of C-13 labelled methane and tiles coated with a low-Z and high-Z marker layer, were used to enhance the volume of information on the material transport. The results show significant asymmetry in the distribution of fuel and plasma impurity species between the inner (net deposition area) and the outer (net erosion) divertor channels. No significant formation of highly hydrogenated carbon films has been found in the Gas Box structure. The important processes for material migration, and the influence of operation scenarios on the morphology of the deposits are discussed. Comparison is also made with results obtained following previous divertor campaigns.
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| 19. |
- Coad, J. P., et al.
(författare)
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Surface analysis of tiles and samples exposed to the first JET campaigns with the ITER-like wall
- 2014
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Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T159, s. 014012-
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Tidskriftsartikel (refereegranskat)abstract
- This paper reports on the first post-mortem analyses of tiles removed from JET after the first campaigns with the ITER-like wall (ILW) during 2011-12 [1]. Tiles from the divertor have been analysed by ion beam analysis techniques and by secondary ion mass spectrometry to determine the amount of beryllium deposition and deuterium retention in the tiles exposed to the scrape-off layer. Films 10-20 mu m thick were present at the top of tile 1, but only very thin films (<1 mu m) were found in the shadowed areas and on other divertor tiles. The total amount of Be found in the divertor following the ILW campaign was a factor of similar to 9 less than the material deposited in the 2007-09 carbon campaign, after allowing for the longer operations in 2007-09.
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| 20. |
- Coenen, J. W., et al.
(författare)
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ELM-induced transient tungsten melting in the JET divertor
- 2015
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:2
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Tidskriftsartikel (refereegranskat)abstract
- The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of I-P = 3.0 MA/B-T = 2.9 T H-mode pulses with an input power of P-IN = 23 MW, a stored energy of similar to 6 MJ and regular type I ELMs at Delta W-ELM = 0.3 MJ and f(ELM) similar to 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within similar to 1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (delta W similar to 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (similar to 80 mu m) were released. Almost 1 mm (similar to 6 mm(3)) of W was moved by similar to 150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j x B forces. The evaporation rate determined from spectroscopy is 100 times less than expected from steady state melting and is thus consistent only with transient melting during the individual ELMs. Analysis of IR data and spectroscopy together with modelling using the MEMOS code Bazylev et al 2009 J. Nucl. Mater. 390-391 810-13 point to transient melting as the main process. 3D MEMOS simulations on the consequences of multiple ELMs on damage of tungsten castellated armour have been performed. These experiments provide the first experimental evidence for the absence of significant melt splashing at transient events resembling mitigated ELMs on ITER and establish a key experimental benchmark for the MEMOS code.
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| 21. |
- Coenen, J. W., et al.
(författare)
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ELM induced tungsten melting and its impact on tokamak operation
- 2015
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 463, s. 78-84
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Tidskriftsartikel (refereegranskat)abstract
- In JET-ILW dedicated melt exposures were performed using a sequence of 3MA/2.9T H-Mode JET pulses with an input power of P-IN = 23 MW, a stored energy of similar to 6 MJ and regular type I ELMs at Delta W-ELM = 0.3 MJ and f(ELM) similar to 30 Hz. In order to assess the risk of starting ITER operations with a full W divertor, one of the task was to measure the consequences of W transients melting due to ELMs. JET is the only tokamak able to produce transients/ ELMs large enough (>300 kJ per ELM) to facilitate melting of tungsten. Such ELMs are comparable to mitigated ELMs expected in ITER. By moving the outer strike point (OSP) onto a dedicated leading edge the base temperature was raised within similar to 1 s to allow transient ELM-driven melting during the subsequent 0.5 s. Almost 1 mm (similar to 6 mm(3)) of W was moved by similar to 150 ELMs within 5 subsequent discharges. Significant material losses in terms of ejections into the plasma were not observed. There is indirect evidence that some small droplets (similar to 80 mu m) were ejected. The impact on the main plasma parameters is minor and no disruptions occurred. The W-melt gradually moved along the lamella edge towards the high field side, driven by j x B forces. The evaporation rate determined is 100 times less than expected from steady state melting and thus only consistent with transient melting during individual ELMs. IR data, spectroscopy, as well as melt modeling point to transient melting. Although the type of damage studied in these JET experiments is unlikely to be experienced in ITER, the results do strongly support the design strategy to avoid exposed edges in the ITER divertor. The JET experiments required a surface at normal incidence and considerable pre-heating to produce tungsten melting. They provide unique experimental evidence for the absence of significant melt splashing at events resembling mitigated ELMs on ITER and establish a unique experimental benchmark for the simulations being used to study transient shallow melting on ITER W divertor PFUs.
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| 22. |
- Coenen, J. W., et al.
(författare)
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Transient induced tungsten melting at the Joint European Torus (JET)
- 2017
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Ingår i: Physica Scripta. - : IOP PUBLISHING LTD. - 0031-8949 .- 1402-4896. ; T170
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Tidskriftsartikel (refereegranskat)abstract
- Melting is one of the major risks associated with tungsten (W) plasma-facing components (PFCs) in tokamaks like JET or ITER. These components are designed such that leading edges and hence excessive plasma heat loads deposited at near normal incidence are avoided. Due to the high stored energies in ITER discharges, shallow surface melting can occur under insufficiently mitigated plasma disruption and so-called edge localised modes-power load transients. A dedicated program was carried out at the JET to study the physics and consequences of W transient melting. Following initial exposures in 2013 (ILW-1) of a W-lamella with leading edge, new experiments have been performed on a sloped surface (15 degrees slope) during the 2015/2016 (ILW-3) campaign. This new experiment allows significantly improved infrared thermography measurements and thus resolved important issue of power loading in the context of the previous leading edge exposures. The new lamella was monitored by local diagnostics: spectroscopy, thermography and high-resolution photography in between discharges. No impact on the main plasma was observed despite a strong increase of the local W source consistent with evaporation. In contrast to the earlier exposure, no droplet emission was observed from the sloped surface. Topological modifications resulting from the melting are clearly visible between discharges on the photographic images. Melt damage can be clearly linked to the infrared measurements: the emissivity drops in zones where melting occurs. In comparison with the previous leading edge experiment, no runaway melt motion is observed, consistent with the hypothesis that the escape of thermionic electrons emitted from the melt zone is largely suppressed in this geometry, where the magnetic field intersects the surface at lower angles than in the case of perpendicular impact on a leading edge. Utilising both exposures allows us to further test the model of the forces driving melt motion that successfully reproduced the findings from the original leading edge exposure. Since the ILW-1 experiments, the exposed misaligned lamella has now been retrieved from the JET machine and post mortem analysis has been performed. No obvious mass loss is observed. Profilometry of the ILW-1 lamella shows the structure of the melt damage which is in line with the modell predictions thus allowing further model validation. Nuclear reaction analysis shows a tenfold reduction in surface deuterium concentration in the molten surface in comparison to the non-molten part of the lamella.
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| 23. |
- Dejarnac, R., et al.
(författare)
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Numerical evaluation of heat flux and surface temperature on a misaligned JET divertor W lamella during ELMs
- 2014
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 54:12, s. 123011-
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Tidskriftsartikel (refereegranskat)abstract
- A series of experiments has been performed on JET to investigate the dynamics of transient melting due to edge localized modes (ELMs). The experiment employs a deliberately misaligned lamella in one module of the JET bulk tungsten outer divertor, allowing the combination of stationary power flux and ELMs to transiently melt the misaligned edge. During the design of the experiment a number of calculations were performed using 2D particle-in-cell simulations and a heat transfer code to investigate the influence on the deposited power flux of finite Larmor radius effects associated with the energetic ELM ions. This has been performed using parameter scans inside a range of pedestal temperatures and densities to scope different experimentally expected ELM energies. On the one hand, we observe optimistic results, with smoothing of the heat flux due to the Larmor gyration on the protruding side of the lamella which sees the direct parallel flux-the deposited power tends to be lower than the nominal value expected from geometric magnetic field line impact over a distance smaller than 2 Larmor radii, a finding which is always valid during ELMs for such a geometry. On the other hand, the fraction of the flux not reaching the directly wetted side is transferred and spread to the top surface of the lamella. The hottest point of the lamella (corner side/top) does not always benefit from the gain from the Larmor smoothing effect because of an enhanced power deposition from the second contribution.
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| 24. |
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| 25. |
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| 26. |
- Esser, H. G., et al.
(författare)
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Material deposition on inner divertor quartz-micro balances during ITER-like wall operation in JET
- 2015
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 463, s. 796-799
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Tidskriftsartikel (refereegranskat)abstract
- The migration of beryllium, tungsten and carbon to remote areas of the inner JET-ILW divertor and the accompanying co-deposition of deuterium has been investigated using post-mortem analysis of the housings of quartz-micro balances (QMBs) and their quartz crystals. The analysis of the deposition provides that the rate of beryllium atoms is significantly reduced compared to the analogue deposition rate of carbon during the carbon wall conditions (JET-C) at the same locations of the QMBs. A reduction factor of 50 was found at the entrance gap to the cryo-pumps while it was 14 under tile 5, the semi-horizontal target plate. The deposits consist of C/Be atomic ratios of typically 0.1-0.5 showing an enrichment of carbon in remote areas compared to directly exposed areas with less carbon. The deuterium retention fraction D/Be is between 0.3 and 1 at these unheated locations in the divertor.
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| 27. |
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| 28. |
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| 29. |
- Heinola, K., et al.
(författare)
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Fuel retention in JET ITER-Like Wall from post-mortem analysis
- 2015
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 463, s. 961-965
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Tidskriftsartikel (refereegranskat)abstract
- Selected Ion Beam Analysis techniques applicable for detecting deuterium and heavier impurities have been used in the post-mortem analyses of tiles removed after the first JET ITER-Like Wall (JET-ILW) campaign. Over half of the retained fuel was measured in the divertor region. The highest figures for fuel retention were obtained from regions with the thickest deposited layers, i.e. in the inner divertor on top of tile 1 and on the High Field Gap Closure tile, which resides deep in the plasma scrape-off layer. Least retention was found in the main chamber high erosion regions, i.e. in the mid-plane of Inner Wall Guard Limiter. The fuel retention values found typically varied with deposition layer thicknesses. The reported retention values support the observed decrease in fuel retention obtained with gas balance experiments of JET-ILW.
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| 30. |
- Heinola, K., et al.
(författare)
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Long-term fuel retention in JET ITER-like wall
- 2016
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Ingår i: Physica Scripta. - : Institute of Physics Publishing (IOPP). - 0031-8949 .- 1402-4896. ; T167
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Tidskriftsartikel (refereegranskat)abstract
- Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITERlike wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be >0.3 mg.
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| 31. |
- Hirai, T., et al.
(författare)
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Characterization and heat flux testing of beryllium coatings on Inconel for JET ITER-like wall project
- 2007
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Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 166-170
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Tidskriftsartikel (refereegranskat)abstract
- In order to perform a fully integrated material test, JET has launched the ITER-like wall project with the aim of installing a full metal wall during the next major shutdown. The material foreseen for the main chamber wall is bulk Be at the limiters and Be coatings on inconel tiles elsewhere. R&D process comprises global characterization ( structure, purity etc) of the evaporated films and testing of their performance under heat loads. The major results are (i) the layers have survived energy loads of 20 MJ m(-2) which is significantly above the required level of 5 - 10 MJ m(-2), (ii) melting limit of beryllium coating would be at the energy level of 30 MJ m(-2), (iii) cyclic thermal load of 10 MJ m(-2) for up to 50 cycles have not induced any noticeable damage such as flaking or detachment.
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| 32. |
- Hirai, T., et al.
(författare)
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R&D on full tungsten divertor and beryllium wall for JET ITER-like wall project
- 2007
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 82:15-24, s. 1839-1845
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Tidskriftsartikel (refereegranskat)abstract
- The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. In order to perform a fully integrated material test JET has launched the ITER-like Wall Project with the aim of installing a full metal wall during the next major shutdown. As a result of R&D projects in 2005-2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies.
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| 33. |
- Hirai, T., et al.
(författare)
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Thermal load testing of erosion-monitoring beryllium marker tile for the ITER-Like Wall Project at JET
- 2008
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 83:7-9, s. 1072-1076
-
Tidskriftsartikel (refereegranskat)abstract
- ITER-Like Wall Project has been launched at JET in order to perform a fully integrated test of plasma-facing materials. During the next major shutdown a full metal wall will be installed: tungsten in the divertor and beryllium in the main chamber. Beryllium erosion is one of key issues to be addressed. Special marker tiles have been designed for this purpose. Test coupons of such markers have been manufactured and examined. The performance test under high power deposition was carried in the electron beam facility JUDITH. The results of material characterization before and after high heat flux loads are presented. The samples survived, without macroscopic damage, power loads of up to 4.5 MW/m(2) for 10s (surface temperature similar to 650 degrees C) and 50 cyclic loads at 3.5 MW/m(2) lasting 10s each (surface temperature similar to 600 degrees C).
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| 34. |
- Horvath, L., et al.
(författare)
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Isotope dependence of the type I ELMy H-mode pedestal in JET-ILW hydrogen and deuterium plasmas
- 2021
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 61:4
-
Tidskriftsartikel (refereegranskat)abstract
- The pedestal structure, edge transport and linear MHD stability have been analyzed in a series of JET with the ITER-like wall hydrogen (H) and deuterium (D) type I ELMy H-mode plasmas. The pedestal pressure is typically higher in D than in H at the same input power and gas rate, with the difference mainly due to lower density in H than in D (Maggi et al (JET Contributors) 2018 Plasma Phys. Control. Fusion 60 014045). A power balance analysis of the pedestal has shown that higher inter-ELM separatrix loss power is required in H than in D to maintain a similar pedestal top pressure. This is qualitatively consistent with a set of interpretative EDGE2D-EIRENE simulations for H and D plasmas, showing that higher edge particle and heat transport coefficients are needed in H than in D to match the experimental profiles. It has also been concluded that the difference in neutral penetration between H and D leads only to minor changes in the upstream density profiles and with trends opposite to experimental observations. This implies that neutral penetration has a minor role in setting the difference between H and D pedestals, but higher ELM and/or inter-ELM transport are likely to be the main players. The interpretative EDGE2D-EIRENE simulations, with simultaneous upstream and outer divertor target profile constraints, have indicated higher separatrix electron temperature in H than in D for a pair of discharges at low fueling gas rate and similar stored energy (which required higher input power in H than in D at the same gas rate). The isotope dependence of linear MHD pedestal stability has been found to be small, but if a higher separatrix temperature is considered in H than in D, this could lead to destabilization of peeling-ballooning modes and shrinking of the stability boundary, qualitatively consistent with the reduced pedestal confinement in H.
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| 35. |
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| 36. |
- 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
-
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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| 37. |
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| 38. |
- Jepu, I, et al.
(författare)
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Beryllium melting and erosion on the upper dump plates in JET during three ITER-like wall campaigns
- 2019
-
Ingår i: Nuclear Fusion. - : Institute of Physics (IOP). - 0029-5515 .- 1741-4326. ; 59:8
-
Tidskriftsartikel (refereegranskat)abstract
- Data on erosion and melting of beryllium upper limiter tiles, so-called dump plates (DP), are presented for all three campaigns in the JET tokamak with the ITER-like wall. High-resolution images of the upper wall of JET show clear signs of flash melting on the ridge of the roofshaped tiles. The melt layers move in the poloidal direction from the inboard to the outboard tile, ending on the last DP tile with an upward going waterfall-like melt structure. Melting was caused mainly by unmitigated plasma disruptions. During three ILW campaigns, around 15% of all 12376 plasma pulses were catalogued as disruptions. Thermocouple data from the upper dump plates tiles showed a reduction in energy delivered by disruptions with fewer extreme events in the third campaign, ILW-3, in comparison to ILW-1 and ILW-2. The total Be erosion assessed via precision weighing of tiles retrieved from JET during shutdowns indicated the increasing mass loss across campaigns of up to 0.6 g from a single tile. The mass of splashed melted Be on the upper walls was also estimated using the high-resolution images of wall components taken after each campaign. The results agree with the total material loss estimated by tile weighing (similar to 130 g). Morphological and structural analysis performed on Be melt layers revealed a multilayer structure of re-solidified material composed mainly of Be and BeO with some heavy metal impurities Ni, Fe, W. IBA analysis performed across the affected tile ridge in both poloidal and toroidal direction revealed a low D concentration, in the range 1-4 x 10(17) D atoms cm(-2).
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| 39. |
- Joffrin, E., et al.
(författare)
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Overview of the JET preparation for deuterium-tritium operation with the ITER like-wall
- 2019
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:11
-
Forskningsöversikt (refereegranskat)abstract
- For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D-T mixtures since 1997 and the first ever D-T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D-T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D-T preparation. This intense preparation includes the review of the physics basis for the D-T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D-T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfven eigenmode antennas, neutral gauges, radiation hard imaging systems...) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D-T campaign provides an incomparable source of information and a basis for the future D-T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas.
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| 40. |
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| 41. |
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| 42. |
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| 43. |
- Krasilnikov, A., et al.
(författare)
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Evidence of 9 Be + p nuclear reactions during 2ω CH and hydrogen minority ICRH in JET-ILW hydrogen and deuterium plasmas
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:2
-
Tidskriftsartikel (refereegranskat)abstract
- The intensity of 9Be + p nuclear fusion reactions was experimentally studied during second harmonic (2ω CH) ion-cyclotron resonance heating (ICRH) and further analyzed during fundamental hydrogen minority ICRH of JET-ILW hydrogen and deuterium plasmas. In relatively low-density plasmas with a high ICRH power, a population of fast H+ ions was created and measured by neutral particle analyzers. Primary and secondary nuclear reaction products, due to 9Be + p interaction, were observed with fast ion loss detectors, γ-ray spectrometers and neutron flux monitors and spectrometers. The possibility of using 9Be(p, d)2α and 9Be(p, α)6Li nuclear reactions to create a population of fast alpha particles and study their behaviour in non-active stage of ITER operation is discussed in the paper.
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| 44. |
- Krieger, K., et al.
(författare)
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Investigation of transient melting of tungsten by ELMs in ASDEX Upgrade
- 2017
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Ingår i: Physica Scripta. - : IOP PUBLISHING LTD. - 0031-8949 .- 1402-4896. ; T170
-
Tidskriftsartikel (refereegranskat)abstract
- Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in the tokamak experiment ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the Divertor Manipulator II system. The exposed sample was designed with an elevated sloped surface inclined against the incident magnetic field to increase the projected parallel power flux to a level were transient melting by ELMs would occur. Sample exposure was controlled by moving the outer strike point to the sample location. As extension to previous melt studies in the new experiment both the current flow from the sample to vessel potential and the local surface temperature were measured with sufficient time resolution to resolve individual ELMs. The experiment provided for the first time a direct link of current flow and surface temperature during transient ELM events. This allows to further constrain the MEMOS melt motion code predictions and to improve the validation of its underlying model assumptions. Post exposure ex situ analysis of the retrieved samples confirms the decreased melt motion observed at shallower magnetic field line to surface angles compared to that at leading edges exposed to the parallel power flux.
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| 45. |
- Likonen, J., et al.
(författare)
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Beryllium accumulation at the inner divertor of JET
- 2005
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 337-39:03-jan, s. 60-64
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Tidskriftsartikel (refereegranskat)abstract
- MkIIGB divertor tiles exposed in JET for the 1998-2001 and 1999-2001 campaigns have been used to assess the amount of beryllium and carbon deposited at the inner divertor wall. Total amount of Be at the inner divertor tiles was determined and integrated toroidally. Results were compared with data obtained with optical spectroscopy and good agreement was obtained. The amount of deposited C was computed from the amount of deposited Be assuming that the Be/C ratio arriving in the divertor is the same as the Be/C ratio in the main chamber. On the basis of this analysis we would expect there to be similar to 0.4 kg of C deposited. This gives an average C deposition rate lower than during the MkIIA phase.
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| 46. |
- Likonen, J., et al.
(författare)
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First results and surface analysis strategy for plasma-facing components after JET operation with the ITER-like wall
- 2014
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Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T159, s. 014016-
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Tidskriftsartikel (refereegranskat)abstract
- During the carbon wall operations of JET since 2001, an extensive post-mortem analysis programme has been carried out under the JET Task Force Fusion Technology and a similar analysis programme is underway for the JET-ILW tiles removed during the 2012 shutdown. The first post-mortem results from the JET ITER-like wall tiles have shown that the overall amount of deposition on the divertor tiles and on remote divertor areas has been reduced by more than an order of magnitude with respect to JET-C. In addition, the obtained data indicate a possible interaction between Be and W such as the formation of mixed Be-W layers. This could be due to the surface roughness of the tiles, or could be caused by diffusion or even alloying. Ion-beam analyses and secondary ion mass spectrometry techniques give only elemental information, so other techniques such as x-ray diffraction, x-ray photoelectron spectroscopy, secondary electron microscopy/energy dispersive x-ray spectroscopy and nuclear microprobing are required. Since the nature of deposition and erosion has changed during the JET-ILW operations, a change in the post-mortem analysis programme is needed. For example, no cross-sectional samples from the sloping parts of tiles 4 and 6 are required. A strategy for post-mortem analyses of the marker-coated tiles will be presented in this paper.
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| 47. |
- Litaudon, X., et al.
(författare)
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Development of steady-state scenarios compatible with ITER-like wall conditions
- 2007
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Ingår i: Plasma Physics and Controlled Fusion. - 0741-3335 .- 1361-6587. ; 49:12B, s. B529-B550
-
Tidskriftsartikel (refereegranskat)abstract
- A key issue for steady-state tokamak operation is to determine the edge conditions that are compatible both with good core confinement and with the power handling and plasma exhaust capabilities of the plasma facing components (PFCs) and divertor systems. A quantitative response to this open question will provide a robust scientific basis for reliable extrapolation of present regimes to an ITER compatible steady-state scenario. In this context, the JET programme addressing steady-state operation is focused on the development of non-inductive, high confinement plasmas with the constraints imposed by the PFCs. A new beryllium main chamber wall and tungsten divertor together with an upgrade of the heating/fuelling capability are currently in preparation at JET. Operation at higher power with this ITER-like wall will impose new constraints on non-inductive scenarios. Recent experiments have focused on the preparation for this new phase of JET operation. In this paper, progress in the development of advanced tokamak (AT) scenarios at JET is reviewed keeping this long-term objective in mind. The approach has consisted of addressing various critical issues separately during the 2006-2007 campaigns with a view to full scenario integration when the JET upgrades are complete. Regimes with internal transport barriers (ITBs) have been developed at q(95) similar to 5 and high triangularity, 3 (relevant to the ITER steady-state demonstration) by applying more than 30 MW of additional heating power reaching beta(N) similar to 2 at B(o) similar to 3.1 T. Operating at higher 6 has allowed the edge pedestal and core densities to be increased pushing the ion temperature closer to that of the electrons. Although not yet fully integrated into a performance enhancing ITB scenario, Neon seeding has been successfully explored to increase the radiated power fraction (up to 60%), providing significant reduction of target tile power fluxes (and hence temperatures) and mitigation of edge localized mode (ELM) activity. At reduced toroidal magnetic field strength, high beta(N) regimes have been achieved and q-profile optimization investigated for use in steady-state scenarios. Values of beta(N) above the 'no-wall magnetohydrodynamic limit' (beta(N) similar to 3.0) have been sustained for a resistive current diffusion time in high-delta configurations (at 1.2 MA/1.8 T). In this scenario, ELM activity has been mitigated by applying magnetic perturbations using error field correction coils to provide ergodization of the magnetic field at the plasma edge. In a highly shaped, quasi-double null X-point configuration, ITBs have been generated on the ion heat transport channel and combined with 'grassy' ELMs with similar to 30 MW of applied heating power (at 1.2 MA/2.7 T, q(95) similar to 7). Advanced algorithms and system identification procedures have been developed with a view to developing simultaneously temperature and q-profile control in real-time. These techniques have so far been applied to the control of the q-profile evolution in JET AT scenarios.
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| 48. |
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| 49. |
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| 50. |
- Lungu, C. P., et al.
(författare)
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Beryllium coatings on metals for marker tiles at JET : development of process and characterization of layers
- 2007
-
Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 157-161
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Tidskriftsartikel (refereegranskat)abstract
- Preparatory study for the operation of the JET tokamak with a full metal wall (ITER-like wall project) also comprises several activities aiming at the development of thin beryllium coatings. The purpose is 2-fold: ( i) to coat Inconel (R) tiles of the inner wall cladding; (ii) to develop methods for production of films for so-called marker tiles in order to enable monitoring of Be erosion from limiters. Properties of the marker film must match, as closely as possible, those of bulk Be. The first step in the R&D process was to assess coating methods and the quality of layers deposited on test coupons. Smooth, dense Be films of high purity and good adhesion to the substrate were deposited with an average deposition rate of 5 +/- 0.5 nm s(-1) to a thickness of 7.5 mu m. A marker structure consisting of a 7.5 mu m Be film on top of a 2.5 mu m Ni interlayer deposited on a bulk Be block has been developed and characterized by means of material analysis methods. An overview of manufacturing processes and properties of the marker coatings is presented.
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