A.P. Zakharov

Hydrogen adsorption on and solubility in graphites

The experimental data on sorption and solubility of hydrogen isotopes in graphite in a wide ranges of temperature and pressure are reviewed. The Langmuir type adsorption is proposed for the hydrogen -- graphites interaction with taking into account dangling sp{sup 2}{minus}bonds relaxation. Three kinds of traps are proposed: Carbon interstitial loops with the adsorption enthalpy of {minus}4.4 eV/H{sub 2} (Traps l); carbon network edge atoms with the adsorption enthalpy of {minus}2.3 eV/H{sub 2} (Traps 2): Basal planes adsorption sites with enthalpy of +2.43 eV/H{sub 2} (Traps 3). The sorption capacity of every kind of graphite could be described with its own unique set of traps. The number of potential sites for the ``true solubility`` (Traps 3) we assume as 1E+6 appm, or HC=l, but endothermic character of this solubility leads to negligible amount of inventory in comparison with Traps 1 and Traps 2. The irradiation with neutrons or carbon atoms increases the number of Traps 1 and Traps 2. At damage level of {approximately}1 dpa under room temperature irradiation the number of these traps was increased up to 1500 and 5000 appm respectively. Traps 1 and Traps 2 are stable under high temperature annealing.

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.

Progress in the development of deposition prevention and cleaning techniques of in-vessel optics in ITER

The lifetime of front optical components unprotected from reactor grade plasmas may be very short due to intensive contamination with carbon and beryllium-based materials eroded by the plasma from beryllium walls and carbon tiles. Deposits result in a significant reduction and spectral alterations of optical transmission. In addition, even rather thin and transparent deposits can dramatically change the shape of reflectance spectra, especially for mirrors with rather low reflectivity, such as W or Mo. The distortion of data obtained with various optical diagnostics may affect the safe operation of ITER. Therefore, the development of optics-cleaning and deposition-mitigating techniques is a key factor in the construction and operation of optical diagnostics in ITER. The problem is of particular concern for optical elements positioned in the divertor region. The latest achievements in protection of in-vessel optics are presented using the example of deposition prevention/cleaning techniques for in-machine components of the Thomson scattering system in the divertor. Careful consideration of well-known and novel protection approaches shows that neither of them alone provides guaranteed survivability of the first in-vessel optics in the divertor. Only a set of complementary prevention/cleaning techniques, which include special materials for mirrors and inhibition additives for plasma, is able to manage the challenging task. The essential issue, which needs to be addressed in the immediate future, is an extensive development of techniques tested under experimental conditions (exposure time and contamination fluxes) similar to those expected in ITER.

First mirrors in ITER: material choice and deposition prevention/cleaning techniques

We present here our recent results on the development and testing of the first mirrors for the divertor Thomson scattering diagnostics in ITER. The Thomson scattering system is based on several large-scale (tens of centimetres) mirrors that will be located in an area with extremely high (3?10%) concentration of contaminants (mainly hydrocarbons) and our main concern is to prevent deposition-induced loss of mirror reflectivity in the spectral range 1000?1064?nm. The suggested design of the mirrors?a high-reflective metal layer on a Si substrate with an oxide coating?combines highly stable optical characteristics under deposition-dominated conditions with excellent mechanical properties. For the mirror layer materials we consider Ag and Al allowing the possibility of sharing the Thomson scattering mirror collecting system with a laser-induced fluorescence system operating in the visible range. Neutron tests of the mirrors of this design are presented along with numerical simulation of radiation damage and transmutation of mirror materials. To provide active protection of the large-scale mirrors we use a number of deposition-mitigating techniques simultaneously. Two main techniques among them, plasma treatment and blowing-out, are considered in detail. The plasma conditions appropriate for mirror cleaning are determined from experiments using plasma-induced erosion/deposition in a CH4/H2 gas mixture. We also report data on the numerical simulation of plasma parameters of a capacitively-coupled discharge calculated using a commercial CFD-ACE code. A comparison of these data with the results for mirror testing under deuterium ion bombardment illustrates the possibility of using the capacitively-coupled discharge for in situ non-destructive deposition mitigation/cleaning.

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.

Characterization of a-B/C:H films deposited from different boron containing precursors

This paper describes the characteristics of a-B/C:H films deposited by PCVD processes with a new harmless precursor-carborane (C2B10H12) in the model device as well as in the T-11M tokamak. Deposition conditions, density, thickness, structure, chemical bonds, content and depth distribution of elements, and some other properties of the films, were investigated in detail. The structure and properties of the films were compared with those for films deposited by other boronization techniques with different precursors (diborane and trimethylborane) and also with a-C:H films. Application of carborane for boronization leads us to the increase of B/C ratio in the films up to a value of 3–4 which is optimal for the erosion process. The structure of the nearest order became icosahedral instead of a mixture of diamond-like and icosahedral as in the case of deposition with diborane. All amorphous a-B/C:H films are homogeneous, and uniform in depth. The hydrogen content is the same as for other B containing films.

Heat load material studies: Simulated tokamak disruptions

It is clear that an improved understanding of the effects of tokamak disruptions on plasma facing component materials is needed for the ITER program. Very large energy fluxes are predicted to be deposited in ITER and could be very damaging to the machine. During 1991, Sandia National Laboratories and the University of New Mexico conducted cooperative tokamak disruption simulation experiments at several Soviet facilities. These facilities were located at the Efremov Institute in Leningrad, the Kurchatov Atomic Energy Institute (Troisk and Moscow) and the Institute for Physicial Chemistry of the Soviet Academy of Sciences in Moscow. Erosion of graphite from plasma stream impact is seen to be much less than that observed with laser or electron beams with similar energy fluxes. This, along with other data obtained, seem to suggest that the “vapor shielding” effect is a very important phenomenon in the study of graphite erosion during tokamak disruption.

Boronization of Russian tokamaks from carborane precursors

Abstract A new and cheap boronization technique using the nontoxic and nonexplosive solid substance carborane has been developed and succesfully applied to the Russian tokamaks T-11M, T-3M, T-10 and TUMAN-3. The glow discharge in a mixture of He and carborane vapor produced the amorphous B C coating with the B C ratio varied from 2.0–3.7. The deposition rate was about 150 nm/h. The primary effect of boronization was a significant reduction of the impurity influx and the plasma impurity contamination, a sharp decrease of the plasma radiated power, and a decrease of the effective charge. Boronization strongly suppressed the impurity influx caused by additional plasma heating. ECR- and ICR-heating as well as ECR current drive were more effective in boronized vessels. Boronization resulted in a signficant extension of the Ne- and q-region of stable tokamak operation. The density limit rose strongly. In Ohmic H-mode energy confinement time increased significantly (by a factor of 2) after boronization. It rose linearly with plasma current Ip and was 10 times higher than Neo-Alcator time at maximum current.