Influence of MeV helium implantation on deuterium retention in self-ion implanted tungsten

Markina, E.; Mayer, M.; Elgeti, S.; Schwarz-Selinger, T.

doi:10.1088/0031-8949/2014/t159/014045

Cited by 10 publications

(9 citation statements)

References 16 publications

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“…Therefore, He atoms can stay in W up to very high temperatures. Helium trapped in single vacancies, vacancy clusters, and bubbles desorbs in the range of 1500-2000 K [14] or even at higher temperatures in the case of the high damage produced by MeV ions [15,16]. Therefore, one could expect a high He concentration in the surface layer under intensive plasma irradiation.…”

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confidence: 99%

Helium concentration measurement in tungsten fuzz-like nanostructures by means of thermal desorption spectroscopy

2016

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The concentration of helium in tungsten fuzz-like nanostructures has been measured by means of thermal desorption spectroscopy. Fuzz was formed on the W surface under intensive plasma irradiation at 1500 K. The helium content was measured first in the as-irradiated sample, and then in a similar sample with the fuzz mechanically scraped from the sample surface. The difference gave the He content in the fuzz, which was estimated to be He/ W = (13 ± 4)%.

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confidence: 99%

Helium concentration measurement in tungsten fuzz-like nanostructures by means of thermal desorption spectroscopy

2016

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“…We investigate how various He concen-trations affect D retention. In a previous study it was shown that 1000 appm as implanted He at 300 K does not affect D retention in self-damaged W [25]. In the experiment presented herein, this finding will be tested for He concentrations larger by one order of magnitude.…”

Section: Introductionmentioning

confidence: 80%

“…Little is known about D retention in He implanted and annealed tungsten. It was shown in literature that when one anneals a He containing W sample no He is released after thermal treatment up to 2000 K [25,30]. This is explained by the high detrapping energy that is necessary to release He from a vacancy or vacancy cluster to which it gets bound during the implantation [9,49].…”

Section: Retention In 1700 K Annealed He Samplesmentioning

confidence: 99%

“…In order to isolate the effect the presence of He has from the effect the creation of displacement damage we have annealed two '3.4 at.% He + W dam' and two 6.8 at.% He containing samples to 1700 K for 30 min. At such high temperatures it was shown that displacement damage is annealed out [48] whereas He stays in the bulk of W [25,30]. Chernikov et al [50] showed that at such annealing temperatures (above 1523 K) visible He bubbles are created in a W sample He irradiated to 600 appm.…”

Section: K Annealed Samplesmentioning

confidence: 99%

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Deuterium transport and retention in the bulk of tungsten containing helium: the effect of helium concentration and microstructure

Markelj¹,

Schwarz-Selinger

Pečovnik³

et al. 2020

Nucl. Fusion

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The effect of helium (He) on deuterium (D) retention and transport in the bulk of tungsten (W) was investigated. For this purpose samples were irradiated by 500 keV He ions at 300 K to different fluences in order to obtain He maximum concentrations of 1 at.%, 3.4 at.%, and 6.8 at.% in 0.84 µm depth. In order to discern the effect of irradiation damage caused by He implantation from the effect of the pure He presence on D retention, the W samples were irradiated at 300 K by 20 MeV tungsten ions in advance to create displacement damage in the crystal lattice to a damage dose of 0.23 dpa. The samples were exposed to a D atom beam at 600 K with a flux of 3.5 ×10 18 D/m 2 s to populate all the created defects. The D depth profiles were measured in situ during and at the end of exposure by nuclear reaction analysis to follow the dynamics of the D uptake. Thermal desorption spectra were collected ex situ at the end of the exposure. We show that D retention increases with implanted He fluence linearly following a D/He ratio of 0.29. We obtained peaking of D concentration at the position of maximum He concentration, reaching for the 6.8 at.% He sample three times higher D concentration (1.1 at. %) than obtained on high dpa W ion irradiated samples (0.37 at. %) for the same loading conditions. D retention and transport was also studied on He containing samples that were annealed to 1700 K. There was no reduction of D retention in the He zone but 80 % reduction in the only W irradiated zone was observed, meaning displacement damage was almost completely removed. In the He zone the D concentration increased to 1.35 at.% and this we attribute to trapping of D around He bubbles of 1.5 nm size created at the He peak maximum as obtained by transmission electron microscopy. Neither in the as He implanted nor the 1700 K annealed sample He did act as diffusion barrier. From this study one can conclude that in the main wall of a future fusion device the effect of He will not dominate D retention in tungsten but at high heat flux areas where displacement damage possibly anneals out He could accumulate in the material that it could eventually dominate over the effect of displacement damage.

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“…Charged particles accelerated to high energies present an efficient means for the production of displacement damage for a reasonable experimental time period at the levels being of interest for fusion research. This method is widely used in fusionoriented investigations [7][8][9][10][11][12][13][14]. The method has been also taken in our research and it was developed on the basis of experimental facilities at NRC "Kurchatov Institute".…”

Section: Introductionmentioning

confidence: 99%

Thermonuclear Fusion Reactor Plasma-Facing Materials under Conditions of Ion Irradiation and Plasma Flux

2021

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A review of experimental studies carried out at the NRC “Kurchatov Institute” on plasma-facing thermonuclear fusion reactor materials is presented in the paper. An experimental method was developed to produce high-level radiation damage in materials simulating the neutron effect by surrogate irradiation with high-energy ions. Plasma-surface interaction is investigated on materials irradiated to high levels of radiation damage in high-flux deuterium plasma. The total fluence of accelerated ions (3–30 MeV, 4He2+, 12C3+, 14N3+, protons) on the samples was 1021–1023 m−2. Experiments were carried out on graphite materials, tungsten, and silicon carbide. Samples have been obtained with a primary defect concentration from 0.1 to 100 displacements per atom, which covers the predicted damage for the ITER and DEMO projects. Erosion dynamics of the irradiated materials in steady-state deuterium plasma, changes of the surface microstructure, and deuterium retention were studied using SEM, TEM, ERDA, TDS, and nuclear backscattering techniques. The surface layer of the materials (3 to hundreds µm) was investigated, and it was shown that the changes in the crystal structure, the loss of their symmetry, and diffusion of defects to grain boundaries play an important role. The most significant results are presented in the paper as an overview of our previous work for many years (carbon and tungsten materials) as well as the relatively recent results (silicon carbide).

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Influence of MeV helium implantation on deuterium retention in self-ion implanted tungsten

Cited by 10 publications

References 16 publications

Helium concentration measurement in tungsten fuzz-like nanostructures by means of thermal desorption spectroscopy

Helium concentration measurement in tungsten fuzz-like nanostructures by means of thermal desorption spectroscopy

Deuterium transport and retention in the bulk of tungsten containing helium: the effect of helium concentration and microstructure

Thermonuclear Fusion Reactor Plasma-Facing Materials under Conditions of Ion Irradiation and Plasma Flux

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