Deuterium retention on the tungsten-coated divertor tiles of JET ITER-like wall in 2015–2016 campaign

Lahtinen, A.; Likonen, J.; Koivuranta, S.; Alves, E.; Baron-Wiechec, A.; Catarino, N.; Coad, J. P.; Heinola, K.; Räisänen, J.; Widdowson, A.; Contributors, Jet

doi:10.1016/j.fusengdes.2019.03.081

Cited by 6 publications

(3 citation statements)

References 11 publications

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“…The OES was thereby not observing the Mo, but a corresponding W tile toroidally away from the analysed PMA tile 3 with Mo finish. It shall be noted that PMA confirmed net W erosion at this spatial location place in the second year of operation with the ILW for a comparable integral period of plasma fluence and plasma conditions [42].…”

Section: Post-mortem Analysis: W Net Erosionsupporting

confidence: 52%

Erosion, screening, and migration of tungsten in the JET divertor

et al. 2019

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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: T, plasma current: MA, auxiliary power MW) in one magnetic configuration accumulating 900 s of plasma with particle fluences in the range of 5– 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...

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Section: Post-mortem Analysis: W Net Erosionsupporting

confidence: 52%

Erosion, screening, and migration of tungsten in the JET divertor

et al. 2019

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“…TDS data are obtained from the facility based at CCFE, United Kingdom using a ramp rate of 10 K min −1 to 775°C (Be) or 1000°C (W and CFC), 1 h anneal and non-controlled ramp down [12]. SIMS is performed at VTT, Finland using O +2 [13]. All systems have capability for handling tritium (T) and beryllium (Be) contaminated samples.…”

Section: Methodsmentioning

confidence: 99%

Evaluation of tritium retention in plasma facing components during JET tritium operations

et al. 2021

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An assessment of the tritium (T) inventory in plasma facing components (PFC) during JET T & DT operations is presented based on the most comprehensive ex-situ fuel retention data set on JET PFCs from the 2015-2016 ILW3 operating period is presented. The global fuel retention is 4.19 x 1023 D atoms, 0.19% of injected fuel. The inner divertor remains the region of highest fuel retention (46.5%). The T inventory in PFCs at the end of JET operations is calculated as 7.48 x 1022 atoms and is informative for accountancy, clean-up efficacy and waste liability assessments. The T accumulation rate at the upper inner divertor during JET deuterium-tritium (DT) operations has been used to assess the requirements and frequency of operation of a new laser induced desorption diagnostic to be installed on JET for the final DT experiments in 2023.

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“…Erosion of the beryllium (Be) first wall and the tungsten (W) divertor in ITER followed by transport of eroded species and their redeposition can result in formation of mixed material layers in some areas. Such layers have been found in JET operating with the ITER-like wall [1][2][3][4]. Fuel species (deuterium (D) and tritium (T)) impinging onto growing layers can get implanted and trapped in them [5][6][7].…”

Section: Introductionmentioning

confidence: 99%

Deuterium retention in mixed Be-W-D codeposited layers

et al. 2020

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Mixed beryllium-tungsten-deuterium (Be-W-D) layers (4.4-28.4 at.% W) were codeposited in a magnetron discharge in D 2 -Ar atmosphere at pressures of 0.8 Pa, 2.7 Pa, and 8 Pa and a substrate temperature of 373 ± 15 K. The composition of the layers was determined using nuclear reaction analysis and Rutherford backscattering spectrometry. D trapping states in the layers were examined using thermal desorption spectroscopy. At all pressures used, the D concentration in the layers has a nonmonotonic dependence on the W concentration, with the maximum occurring at 4.4-7.3% W. In this case, the D concentration is about two times greater than that in a Be-D layer; in W-D layers the D concentration is more than one order of magnitude smaller. Increase of pressure during deposition results in an increase of the D concentration in Be-D and Be-W-D (4.4-7.3% W) layers and is linked with the appearance of sharp low-temperature D release peaks near 450 K and 500 K. Increase of pressure also results in steeper decrease of the D concentration with increasing W concentration beyond 7.3%. It is concluded that for the layers deposited at 450 K and above, the effect of the presence of W (4.4-28.4%) on the D retention becomes small.

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Deuterium retention on the tungsten-coated divertor tiles of JET ITER-like wall in 2015–2016 campaign

Cited by 6 publications

References 11 publications

Erosion, screening, and migration of tungsten in the JET divertor

Erosion, screening, and migration of tungsten in the JET divertor

Evaluation of tritium retention in plasma facing components during JET tritium operations

Deuterium retention in mixed Be-W-D codeposited layers

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