V.Kh. Alimov

Hydrogen isotope retention in plasma-facing materials: Review of recent experimental results

Recent data on deuterium retention in carbon fibre composites and tungsten both irradiated with D ions and exposed to D plasmas are presented. Deuterium depth profiles measured up to depths of 7–14 μm allow understanding of the mechanism which is responsible for the hydrogen isotope trapping in these materials. In the CFC materials the amount of retained deuterium increases with the ion fluence at all irradiation temperatures in the range from 323 to 723 K. No saturation is reached as observed in pyrolytic graphite. Depth profiles show that saturation occurs only within a near surface layer corresponding to the ion range. The increase in total retention at near-room temperature is accompanied by an increasing of the long profile tail extending beyond 14 μm with the D concentration of about 10−1 at.% at a depth of 10 μm for fluences above 1024 D m−2. The depth at which deuterium is retained in tungsten (W) can be divided into three zones: (i) the near-surface layer (up to a depth of 0.2–0.5 μm depending on ion energy), (ii) the sub-surface layer (from ~0.5 to ~2 μm), and (iii) the bulk (>5 μm). Low-energy D ion irradiation modifies the W structure to depths of up to about 5 μm, both for W single crystals and polycrystalline W. The high D concentration (0.1–0.3 at.%) at depths of 1–3 μm relates to accumulation of D2 molecules in vacancy clusters and voids. These defects are supposed to be generated due to plastic deformation of the W surface caused by deuterium supersaturation within the near-surface layer.

Recent advances on hydrogen retention in ITER's plasma-facing materials: Beryllium, carbon and tungsten

Abstract Management of tritium inventory remains one of the grand challenges in the development of fusion energy, and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER’s plasma-facing materials – Be, C, and W – and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this paper together with recommendations for ITER. Basic parameters of diffusivity, solubility, and trapping in Be, C, and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping, but long-term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be- and C-containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak, and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.

Subsurface morphology changes due to deuterium bombardment of tungsten

Recrystallized polycrystalline tungsten was exposed to a deuterium plasma beam with high flux (1022u2009Du2009m−2u2009s−1) and low energy (38u2009eVu2009D−1) to fluences up to 1027u2009Du2009m−2. The sample temperature was varied between 320 and 800u2009K. The three-dimensional morphology of blister-like structures and the grain orientation were investigated by scanning electron microscopy combined with focused ion beam cross-sectioning and electron backscattering diffraction. Cracks with distorted areas ( 480u2009K) were observed beneath the surface. The surface blister-like structures and the defects underneath are correlated along crystallographic orientation of the W grains in accordance to the low-indexed slip system {110}111. The defects are mobile and accumulate under deuterium loading. Samples exposed near room temperature do not form such large cavities by subsequent heating up to 1300u2009K. Deuterium bombardment above 700u2009K does not lead to blister-like structures.

Deuterium retention and re-emission from tungsten materials

Deuterium retention in five types of tungsten samples (W single crystal, W produced by hot pressing, chemical vapor and plasma sprayed deposited W coatings) has been investigated during and after implantation with 1.5 keV D-ions at 300, 600, and 900 K by means of re-emission, thermal desorption spectroscopy and nuclear reaction analysis measurements. Deuterium inventory and its spontaneous and thermal release depend strongly on the structure of the material. At 300 K, the saturated amount of D retained in the samples is in the range 1 × 1017 to 4 × 1017 D/cm2. For higher implantation temperature, 900 K, the retained amount is smaller by a factor of about 50 than for 300 K. During implantation D is retained far beyond the implantation zone. An amount of ≅ 3 × 1016 D/cm2 is trapped in the near-surface layer during implantation at 300 K. More than 50% of the implanted D diffuses into the bulk and is captured by lattice imperfections.

Deuterium retention in tungsten exposed to low-energy pure and helium-seeded deuterium plasmas

Influence of helium (He) on the deuterium (D) retention in tungsten (W) under simultaneous He-D plasma exposure was investigated. Bulk polycrystalline tungsten and two W coatings on carbon substrate, namely, plasma-sprayed tungsten and combined magnetron-sputtered and ion implanted tungsten (CMSII-W) were exposed to pure and He-seeded D plasmas generated by electron-cyclotron-resonance plasma source. The D retention in each sample was subsequently analyzed by various methods such as nuclear reaction analysis for the D depth profiling up to 6u2002μm and thermal desorption spectroscopy for the determination of total amount of D retention. It is shown that seeding of helium into D plasma with helium ion flux fraction of 10% reduces the deuterium retention for all tungsten grades but more significant reduction was observed for polycrystalline W and less significant effect was found for W coatings. From the thermal desorption spectroscopy measurements, we conclude that the presence of He modifies the density of ex...

Deuterium retention and lattice damage in tungsten irradiated with D ions

Depth profiles of D atoms and D2 molecules in a W single crystal implanted with 6 keV D ions at 300 K have been determined using secondary ion mass spectrometry (SIMS) and residual gas analysis (RGA) measurements in the course of surface sputtering. Profiles of deuterium and lattice damage in a W single crystal irradiated with 10 keV D ions at 300 K have been investigated by means of nuclear reaction analysis (NRA) and Rutherford backscattering spectrometry combined with ion channelling techniques (RBS/C). There are at least two types of ion-induced defects which are responsible for trapping of deuterium: (i) D2-filled microvoids (deuterium bubbles) localised in the implantation zone; and (ii) dislocations which are distributed from the surface to depths far beyond 1 μm and which capture deuterium in the form of D atoms.

Interaction of hydrogen with radiation defects in metals

The interaction of hydrogen (deuterium) with radiation defects in Mo and Ni was observed by the hydrogen permeation method from a glow discharge plasma for self-interstitial atoms (SIAs) and by depth profiling of implanted gas for vacancies (V). On the basis of the analogies with the free metal surface the model of interaction of hydrogen with vacancy defects is suggested. Using the known values of the adsorption and solution heats, the binding energies of hydrogen with vacancy and vacancy clusters are estimated for some metals (V, Nb, Ta, α-Fe, W, Pd, Cu, Au, Al). Good correlation is observed for calculated and experimental values.

Depth distribution of deuterium atoms and molecules in beryllium implanted with D ions

Abstract In-depth concentration profiles of deuterium atoms and molecules in beryllium implanted with 9 keV D ions to fluences, Φ, in the range from 6 × 10 19 to 9 × 10 22 D / m 2 at temperatures, T irr , of 300 and 700 K have been determined using SIMS and RGA (residual gas analysis) measurements in the course of surface sputtering. The microstructure of implanted specimens was studied by TEM. Implanted deuterium is retained in Be matrix in the form of both D atoms and D 2 molecules. The total amount of gas captured within the sub-surface layer of ∼ 700 nm in thickness as a result of implantation at 300 and 700 K reaches 4 × 10 21 and 1 × 10 21 D / m 2 , correspondingly. The ratio of deuterium quantities retained in the form of atoms and molecules, Q D : Q D 2 , varies from 1:3 for T irr = 300 K to 1:4 for T irr = 700 K . At T irr = 300 K the concentration of D 2 molecules at the depth of the ion mean range reaches its maximum of 4 × 10 27 molecules / m 3 at Φ ≈ 2 × 10 21 D / m 2 . The molecules are present in tiny bubbles which show a tendency toward interconnection at higher fluences. At T irr = 700 K , along with relatively small facetted bubbles (near the very surface), large oblate gas-filled cavities and channels forming extended labyrinths appear and they accumulate most of the injected gas. The maximum D 2 concentration in the latter case is of 1 × 10 27 molecules / m 3 . The high concentration of D atoms in the ion stopping zone after implantation at T irr = 300 and 700 K (about 2 × 10 27 and 1 × 10 27 atoms / m 3 , respectively) is attributed to deuterium (i) trapped in radiation vacancies, (ii) adsorbed on the walls of bubbles and channels and (iii) bonded to/by BeO formed on the surface and present in the form of metallurgical inclusions in the bulk.

Gas swelling and related phenomena in beryllium implanted with deuterium ions

Abstract An extensive TEM study of the microstructure of TIP-30 Be implanted with 3 and 10 keV D ions to fluences, Φ in the range from 3 × 1020 to 8 × 1021D/m2 at temperatures, Tirr = 300, 500 and 700 K has been carried out. Depth distributions of separate D atoms and D2 molecules have been investigated by means of SIMS and RGA methods, correspondingly. D ion irradiation, accompanied by blistering, gives rise to destructions dependent mainly on Tirr. Irradiation at 300 K leads to the formation of tiny D2 bubbles of 1 run in size (reminiscent of He bubbles in Be). At Tirr ≥ 500 K, along with small facetted bubbles, the development of larger oblate cavities occurs accumulating most of injected deuterium and providing for a much higher gas swelling compared to that at 300 K. D (He) ion implantation leads to the enhanced growth of microcrystalline layers of cph-BeO oxide with a microstructure differing from that on the electropolished Be surface. Based on the analysis of experimental data deuterium reemission, thermal desorption and trapping in defects are discussed.

Gas-induced swelling of beryllium implanted with deuterium ions

Abstract An extensive TEM study of the microstructure of Be TIP-30 irradiated with 3 and 10 keV D ions up to fluences, Φ, in the range from 3 × 10 20 to 8 × 10 21 D/m 2 at temperatures, T irr = 300, 500 and 700 K has been carried out. Depth distributions of deuterium in a form of separate D atoms and D 2 molecules have been investigated by means of SIMS (secondary ion mass spectrometry) and RGA (residual gas analysis) methods, correspondingly. D ion implantation is accompanied by blistering and gives rise to processes of gas-induced cavitation which are very sensitive to the irradiation temperature. At T irr = 300 K tiny gas bubbles (about 1 nm in size) pressurized with molecular deuterium are developed with parameters resembling those of helium bubbles in Be. Irradiation at T irr ≥ 500 K leads to the appearance of coarse deuterium-filled cavities which can form in sub-surface layers different kinds of oblate labyrinth structures. Questions of reemission, thermal desorption and trapping of deuterium in Be have been discussed.