Behavior of tungsten with exposure to deuterium plasmas

Four kinds of tungsten (W) materials, i.e. (1) foil of 50 μm thick (f-W), (2) polycrystalline (Pc-W) with grain size of ∼3 μm, (3) recrystallized (Re-W) with grain size of ∼50 μm and (4) vacuum plasma spraying (VPS-W) coatings, were irradiated employing linear plasma generators, with fluxes ?1 × 10²² D/m²/s and energies ?100 eV/D. Scanning electron microscopy (SEM) was used to observe blister formation at the surfaces. The SEM surface morphology and cross section observation indicates that blister formation is related to the microstructure and surface state of different material grades. Results of trapping and deuterium retention measured by thermal desorption spectroscopy (TDS) and nuclear reaction analysis (NRA) show also a close correlation between the retention and the microstructure and surface state.     

Dust particle formation due to interaction between graphite and helicon deuterium plasmas

The collection of dust particles using divertor simulation helicon plasmas has been carried out to examine dust formation due to the interaction between a graphite target and deuterium plasmas, which are planned to operate in the large helical device (LHD) at the Japanese National Institute for Fusion Science (NIFS). The collected dust particles are classified into three types: (i) small spherical particles below 400 nm in size, (ii) agglomerates whose primary particles have a size of about 10 nm, and (iii) large flakes above 1 μm in size. These features are quite similar to those obtained through hydrogen plasma operation, indicating that the dust formation mechanisms due to the interaction between a carbon wall and a plasma of deuterium, which is the isotope of hydrogen, is probably similar to those of hydrogen.     

Optical characteristics of recrystallized tungsten mirrors exposed to low-energy, high flux D plasmas

The surface topography and optical properties of recrystallized tungsten exposed to a low-energy (38 eV/D), high flux (10²² D/m² s) deuterium plasma with an ion fluence of 10²⁶ D/m² at various temperatures was investigated. It was found that the surface morphology weakly depends on the exposure temperature in the range 320-695 K with the exception of the narrow temperature region around 535 K, where large changes to all optical characteristics occurs. After plasma exposure at this temperature, the surface topography of the W sample is characterized by active blistering as has already been indicated in previous publications. The reflectance found in direct measurements at normal incidence drops in the wavelength interval 220-650 nm, whereas the estimations of reflectance using the ellipsometry data demonstrate some increase.     

Numerical estimates of the ITER first mirrors degradation due to atomic fluxes

A systematic numerical study of the deposition of Be and C atoms on the ITER main chamber first mirrors and their erosion due to fast D(T) atoms has been carried out. The calculations are based on Monte-Carlo neutral transport modelling (with the EIRENE code) on a simulated steady-state plasma background. A generic reduced model of cylindrical and conical diagnostic ducts is considered. The results indicate that the sputtering of the mirrors can be made acceptably small if they are installed in sufficiently long ducts. At the same time the deposition of impurities can be a serious issue even for mirrors protected in long channels. However, the calculations of deposition are not very reliable at the current stage and a need to seriously improve them is indicated.     

Saturation in deuterium retention of CFC graphite targets exposed to beryllium-seeded plasmas on PISCES-B

ITER strike-plates are foreseen to be of carbon-fiber-composite (CFC). In this study the CFC bulk deuterium retention in ITER-relevant conditions is investigated. DMS 701 (Dunlop) CFC targets were exposed to plasma in PISCES-B divertor plasma simulator. Samples were exposed to both pure deuterium plasma and beryllium-seeded plasma at high fluences (up to ) and high surface temperature (1070 K). The deuterium contents of the exposed samples have been measured using both thermal-desorption-spectrometry (TDS) during baking at 1400 K and ion beam nuclear reaction analysis (NRA). The total deuterium inventory has been obtained from TDS while NRA measured the deuterium depth distribution. In the analysed fluence range at target temperature of 1070 K, no fluence dependence was observed. The measured released deuterium is . In the case of target exposure with beryllium-seeded plasma no change in the released amount of deuterium was found. The deuterium concentration inside the samples is almost constant until the probed depth of ?m, except in the first 1 μm surface layer, where it is 5 times higher than in the bulk. No C erosion/redeposition was observed in the Be-seeded plasma cases. The measured retention, applied to 50 m² of ITER CFC surface, would imply a tritium saturated value of 0.3 gT, much lower than the ITER safety limit of 350 g.     

The effect of displacement damage on deuterium retention in ITER-grade tungsten exposed to low-energy,high-flux pure and helium-seeded deuterium plasmas

Samples prepared from polycrystalline ITER-grade tungsten were damaged by irradiation with 20 MeV W ions at room temperature to a fluence of 1.4 × 10¹⁸ W/m². Due to the irradiation, displacement damage peaked near the end-of-range, 1.35 μm beneath the surface, at 0.89 displacements per atom. The damaged as well as undamaged W samples were then exposed to low-energy, high-flux (10²² D/m² s) pure D and helium-seeded D plasmas to an ion fluence of 3 × 10²⁶ D/m² at various temperatures. Trapping of deuterium was examined by the D(³He,p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV allowing determination of the D concentration at depths up to 6 μm. It has been found that (i) addition of 10% helium ions into the D plasma at exposure temperatures of 440–650 K significantly reduces the D concentration at depths of 0.5–6 μm compared to that for the pure plasma exposure; (ii) generation of the W-ion-induced displacement damage significantly increases the D concentration at depths up to 2 μm (i.e., in the damage zone) under subsequent exposures to both pure D and D–He plasmas.     

Deuterium retention in tungsten exposed to low-energy, high-flux clean and carbon-seeded deuterium plasmas

Depth profiles of deuterium trapped in tungsten exposed to a low-energy (≈200 eV/D) and high deuterium ion flux (about 1 × 10²¹ D/m² s) in clean (We use the term ‘clean’ in quotation marks having in mind the impossibility to obtain absolutely clean plasma. In our case the conception ‘clean’ D plasma means the plasma without intentionally introduced carbon impurities.) and carbon-seeded D plasmas at an ion fluence of about 2 × 10²⁴ D/m² and various temperatures have been measured up to a depth of 7 μm using the D(³He, p)⁴He nuclear reaction at a ³He energy varied from 0.69 to 4.0 MeV. The deuterium retention in single-crystalline and polycrystalline W increases with the exposure temperature, reaching its maximum value at about 500 K (for ‘clean’ plasma) or about 600 K (for carbon-seeded plasma), and then decreases as the temperature grows further. It is assumed that tungsten carbide formed on the W surface under exposure to the carbon-seeded D plasmas serves as a barrier layer for diffusion and prevents the outward transport of deuterium, thus increasing the D retention in the bulk of tungsten.     

Carbon film growth and hydrogenic retention of tungsten exposed to carbon-seeded high density deuterium plasmas

Tungsten (W) targets have been exposed to high density (n_e ? 4 × 10¹⁹ m^?3), low temperature (T_e ? 3 eV) CH₄-seeded deuterium (D) plasma in Pilot-PSI. The surface temperature of the target was ～1220 K at the center and decreased radially to ～650 K at the edges. Carbon film growth was found to only occur in regions where there was a clear CII emission line, corresponding to regions in the plasma with T_e ? 2 eV. The maximum film thickness was ～2.1 μm after a plasma exposure time of 120 s. ³He nuclear reaction (NRA) analysis and thermal desorption spectroscopy (TDS) determine that the presence of a thin carbon film dominates the hydrogenic retention properties of the W substrate. Thermal desorption spectroscopy analysis shows retention increasing roughly linearly with incident plasma fluence. NRA measures a C/D ratio of ～0.002 in these films deposited at high surface temperatures.     

Optical properties of Al mirrors under impact of deuterium plasma ions in experiments simulating ITER conditions

The ion-induced modification of aluminum alloy mirrors, under bombardment by deuterium plasma ions has been investigated as a simulation of the environment effects on in-vessel mirrors in ITER. Ellipsometry and reflectrometry have been used to characterize the mirror surface, along with several surface diagnostic techniques (XPS, Auger, SIMS). The results of multiangular- and spectro-ellipsometry were analyzed using both a bare surface model, and effective medium model; the medium was composed of Al, Al₂O₃ (Al(OD)₃ or AlOOD), and voids. It was found that the reflectance decreases following exposure to keV-range ions, but can be restored by subsequent exposing the mirror to low-energy ions (∼60 eV). Chemical processes related to an increased oxide layer are thought to be responsible for the decrease in reflectance, while the reduction of the oxide layer following low-energy D⁺ exposure may lead to the return of high reflectance. By comparing the measurements with the results of modeling, a mechanism is suggested to explain the experimental data. The mechanism is based on: (1) chemical processes in a surface layer and (2) сhanges in the thickness and roughness of the surface layer.     

Spectroscopic studies of homogeneous thin carbon erosion films on mirrors and flakes with a high deuterium content formed in tokamak T-10

Redeposited hydrocarbon films on plasma facing elements in tokamaks accumulate hydrogen isotopes. In the present study such films were made to redeposit on stainless steel mirror substrates as thin films and without any substrate as bare flakes with high deuterium content, under deuterium-plasma discharges inside T-10 tokamak vacuum chamber. These films were subjected to spectral characterizations through Fourier-transform infrared (FT-IR), electron paramagnetic resonance (EPR), and photoluminescence techniques. IR spectra showed the presence of two main deuterium states as observed by the CD_2,3 sp³ stretching modes at 2100–2200 cm⁻¹ and the CD₂ sp³ bending modes at 600–1100 cm⁻¹. Among these, CD₃ stretching mode at 2217 cm⁻¹ may serve as a control for deuterium desorption during the cleanup process of the reactor. As a comparative measure, C60 films were also studied, the luminescence excitation spectrum of which showed similarity in peak positions with tokamak bare flakes pertained to sp² luminescence centers. The observed spectral differences are mainly due to more localized sp² states for C60 and sp³ states for tokamak flakes. EPR spectra of the bare flakes showed the defective states with a high spin density, ∼10¹⁹ cm⁻³ which serve as luminescence quenching centers, and provide a path for hydrogen isotopes adsorption.     

Erosion of thin hydrogenated carbon films in oxygen,oxygen/hydrogen and water plasmas

The erosion of amorphous hydrogenated carbon films in oxygen, oxygen/hydrogen and water electron cyclotron resonance plasmas was investigated by in situ ellipsometry. The erosion was measured as a function of the energy of the impinging ions and the substrate temperature. Erosion is most effective in pure oxygen plasmas. The erosion rate rises with increasing ion energy and substrate temperature, in the latter case however only at low ion energies. The reaction layer at the surface of the eroded film is further analyzed by X-ray photoelectron spectroscopy (XPS). The C 1s peak of the XPS spectra shows, that oxygen is implanted in the films and forms double and single bonds to the carbon atoms. This modification, however, is limited to a few atomic layers.     

Study of deuterium retention on lithiated tungsten exposed to high-flux deuterium plasma using laser-induced breakdown spectroscopy

Tungsten is under consideration for use as a plasma-facing material in the divertor region of ITER. Lithiation can significantly improve plasma performance in long-pulse tokamaks like EAST. The investigation of lithiated tungsten is important for understanding the lithium conditioning effects for EAST, where tungsten will be used as a plasma-facing material. In this paper, a few important issues of lithiated tungsten interacting with high-flux deuterium plasma have been studied, such as the effect of lithiation on deuterium retention, the profile of elemental distribution, and the chemical state of lithiated tungsten. Deuterium retention inside both pure and lithiated tungsten has been investigated for the first time in the linear plasma simulator Magnum-PSI by in-situ laser induced breakdown spectroscopy (LIBS). The results indicate that, after deuterium plasma exposure, deuterium retention could be saturated in the lithiation layer, and the lithium in the lithiated layer is chemically bound with deuterium. Moreover, the lithiation can inhibit the blistering on the tungsten surface. These results can be valuable for the application of LIBS as a diagnostic technique for plasma-facing components of tokamaks.     

Interaction of deuterium with lattice defects in nickel

The trapping of ion-implanted deuterium (D) in fee Ni is investigated by ion-beam-analysis techniques. Two lattice-defect traps have been observed with trap-binding enthalpies 0.24 eV and 0.43 eV referred to an untrapped solution site. The lattice location of D when associated with the defect traps is obtained by the channeling technique following anneals at various temperatures. The detailed analysis of these channeling data is based on a comparison with multirow continuum-model calculations of the angular yields for different D positions. These channeling calculations are extended by introducing a parameter δψ which encompasses the spreading in transverse energy caused by effects such as, for example, electron and nuclear multiple scattering. Also new and improved theoretical calculations based on the effective medium scheme of the equilibrium positions of H isotopes at defects, especially vacancies, are presented. The calculations show that D is delocalized over the entire vacancy, with a maximum density in the region between the vacancy and the nearest octahedral site. This picture is supported by the finding that the channeling data for D trapped to vacancies cannot be interpreted in terms of a single lattice site, and preference in site occupancy is found for D displaced from the vacancy towards the octahedral and (smaller) tetrahedral sites, respectively.     

Generalization of Solovev's approach to finding equilibrium solutions for axisymmetric plasmas with flow

Solovev's approach of finding equilibrium solutions was found to be extremely useful for generating a library of linear-superposable equilibria for the purpose of shaping studies.This set of solutions was subsequently expanded to include the vacuum solutions of Zheng,Wootton and Solano,resulting in a set of functions [SOLOVEV_ZWS] that were usually used for all toroidally symmetric plasmas,commonly recognized as being able to accommodate any desired plasma shapes (complete-shaping capability).The possibility of extending the Solovev approach to toroidal equilibria with a general plasma flow is examined theoretically.We found that the only meaningful extension is to plasmas with a pure toroidal rotation and with a constant Mach number.We also show that the simplification ansatz made to the current profiles,which was the basis of the Solovev approach,should be applied more systematically to include an internal boundary condition at the magnetic axis;resulting in a modified and more useful set [SOLOVEV_ZWSm].Explicit expressions of functions in this set are given for equilibria with a quasi-constant current density profile,with a toroidal flow at a constant Mach number and with specific heat capacity 1.The properties of [SOLOVEV_ZWSm] are studied analytically.Numerical examples of achievable equilibria are demonstrated.Although the shaping capability of the set [SOLOVE_ZWSm] is quite extensive,it nevertheless still does not have complete shaping capability,particularly for plasmas with negative curvature points on the plasmaboundary such as the doublets or indented bean shaped tokamaks.