Deuterium retention in plasma spray tungsten coatings exposed to low-energy, high flux D plasma

Two types of porous plasma spray tungsten coatings deposited onto stainless steel and graphite substrates were exposed to low-energy (76 eV ), high-flux (10²² D/m² s) D plasma to ion fluences of (3-4) × 10²⁶ D/m² at various temperatures. Deuterium retention in the W coatings was examined by thermal desorption spectroscopy and the D(³He,p)⁴He nuclear reaction, allowing determination of the D concentration at depths up to 7 μm. The relatively high D concentration (above 0.1 at.%) at depths of several micrometers observed after D plasma exposure at 340-560 K can be related to accumulation of D₂ molecules in pores, while at temperatures above 600 K deuterium is accumulated mainly in the form of D atoms chemisorbed on the inner pore surfaces. At exposure temperatures above 500 K, the D retention in the plasma spray W coating on graphite substrate increases significantly due to trapping of diffusing D atoms at carbon dangling bonds located at the edge of a graphite crystallite.     

Surface morphology and deuterium retention in tungsten oxide layers exposed to low-energy, high flux D plasma

Surface morphology and deuterium retention in tungsten oxide layers (WO_3−z, z ? 0.25) grown on polycrystalline and recrystallized W substrates have been examined after exposure to a low-energy (38 eV/D), high flux (10²² D/m² s) D plasma to an ion fluence of 10²⁶ D/m² at various temperatures (up to ∼700 K). Characterization methods used were scanning electron microscopy, X-ray diffraction, Rutherford backscattering spectroscopy, and the D(³He,p)⁴He nuclear reaction analysis. During exposure to the D plasma at temperatures of 340-615 K, a partial reduction of the tungsten oxide takes place in the near-surface layer up to 0.3 μm in depth. Even at around room temperature, deuterium atoms diffuse several micrometers into the tungsten oxide. The high D concentration of about 0.1 D/W observed in the first micrometers below the surface at temperatures below 500 K can be related mainly to D atoms chemically bonded to O atoms. As the exposure temperature increases, the D concentration decreases, reaching about 2 × 10⁻⁴ D/W at 615 K. At plasma exposure temperatures of about 700 K, the oxide layer shrinks and loses a large fraction of oxygen.     

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.     

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.     

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.     

Qualification and post-mortem characterization of tungsten mock-ups exposed to cyclic high heat flux loading

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, high heat flux tests were performed in the electron beam facility FE200, Le Creusot, France. Thereby, in total eight small-scale and three medium-scale monoblock mock-ups produced with different manufacturing technologies and different tungsten grades were exposed to cyclic steady state heat loads. The applied power density ranges from 10 to 20 MW/m² with a maximum of 1000 cycles at each particular loading step. Finally, on a reduced number of tiles, critical heat flux tests in the range of 30 MW/m² were performed.Besides macroscopic and microscopic images of the loaded surface areas, detailed metallographic analyses were performed in order to characterize the occurring damages, i.e., crack formation, recrystallization, and melting. Thereby, the different joining technologies, i.e., hot radial pressing (HRP) vs. hot isostatic pressing (HIP) of tungsten to the Cu-based cooling tube, were qualified showing a higher stability and reproducibility of the HIP technology also as repair technology. Finally, the material response at the loaded top surface was found to be depending on the material grade, microstructural orientation, and recrystallization state of the material. These damages might be triggered by the application of thermal shock loads during electron beam surface scanning and not by the steady state heat load only. However, the superposition of thermal fatigue loads and thermal shocks as also expected during ELMs in ITER gives a first impression of the possible severe material degradation at the surface during operational scenarios at the divertor strike point.     

Temperature dependence of surface morphology and deuterium retention in polycrystalline ITER-grade tungsten exposed to low-energy,high-flux D plasma

Surface topography and deuterium retention in polycrystalline ITER-grade tungsten have been examined after exposure to a low-energy (38 eV/D), high-flux (10²² D/m² s) deuterium plasma with ion fluences of 10²⁶ and 10²⁷ D/m² at various temperatures. The methods used were scanning electron microscopy equipped with focused ion beam, thermal desorption spectroscopy, and the D(³He,p) ⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 320 to 815 K, small blisters of size in the range from 0.2 to 5 μm, depending on the exposure temperature and ion fluence, are formed on the W surface. At an ion fluence of 10²⁷ D/m², the deuterium retention increases with the exposure temperature, reaching its maximum value of about 10²² D/m² at 500 K, and then decreases below 10¹⁹ D/m² at 800 K.     

Thermal response of plasma sprayed tungsten coating to high heat flux

In order to investigate the thermal response of tungsten coating on carbon and copper substrates by vacuum plasma spray (VPS) or inert gas plasma spray (IPS), annealing and cyclic heat load experiments of these coatings were conducted. It is indicated that the multi-layered tungsten and rhenium interface of VPS-W/CFC failed to act as a diffusion barrier at elevated temperature and tungsten carbides were developed after 1 h incubation time when annealing temperature was higher than 1600 °C. IPS-W/Cu and W/C without an intermediate bonding layer were failed by the detachment of the tungsten coating at 900 and 1200 °C annealing for several hours, respectively. Cyclic heat load of electron beam with 35 MW/m² and 3-s pulse duration indicated that IPS-W/Cu samples failed with local detachment of the tungsten coating within 200 cycles and IPS-W/C showed local cracks by 300 cycles, but VPS-W/CFC withstood 1000 cycles without visible damages. However, crack creation and propagation in VPS-W/CFC were also observed under higher heat load.     

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.     

Changes to the reflectance of Be mirrors due to deuterium plasmas contaminated with oxygen

The reflectance of Be mirrors due to impact by ions from a deuterium plasma has been studied under several bombardment conditions. Analysis of the resulting surface films has been performed using various diagnostic techniques, with the conclusion that the primary factor leading to the decrease in reflectance following bombardment with energetic ions is the conversion of the surface oxide layer, composed of BeO, to the hydroxide, Be(OD)₂, with a corresponding increase in the optical extinction coefficient. The increase in the thickness of the layer is also important. Modifications to the surface layer are thought to involve a balance between the ion-induced diffusion of Be atoms to the surface where they may react with incident D and O atoms, and physical and chemical sputtering processes. For incident ion energies less than ∼50 eV, chemical reactions leading to disoxidation of the oxide-hydroxide film dominate, while keV-range ions (primarily D, but with some O impurities) lead to the formation of hydroxide, and an increase in the surface layer thickness.     

Operational characteristics of the high flux plasma generator Magnum-PSI

In Magnum-PSI (MAgnetized plasma Generator and NUMerical modeling for Plasma Surface Interactions), the high density, low temperature plasma of a wall stabilized dc cascaded arc is confined to a magnetized plasma beam by a quasi-steady state axial magnetic field up to 1.3 T. It aims at conditions that enable fundamental studies of plasma–surface interactions in the regime relevant for fusion reactors such as ITER: 10²³–10²⁵ m⁻² s⁻¹ hydrogen plasma flux densities at 1–5 eV. To study the effects of transient heat loads on a plasma-facing surface, a high power pulsed magnetized arc discharge has been developed. Additionally, the target surface can be transiently heated with a pulsed laser system during plasma exposure. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.     

Influence of high flux hydrogen-plasma exposure on the thermal shock induced crack formation in tungsten

The influence of high flux hydrogen-plasma on the thermal shock behaviour of tungsten was investigated in a combined experiment using the linear plasma device Pilot-PSI and the electron beam facility JUDITH 1. Tungsten targets were exposed to high flux hydrogen plasma, cyclic thermal shock tests and a combination of both loading conditions. The induced thermal shock crack networks and surface modifications were investigated and characterised using scanning electron microscopy (SEM), optical microscopy and laser profilometry. Comparisons of the results showed that the combination of hydrogen plasma and thermal shock loading has a significant influence on the crack pattern in terms of crack distance, width and depth. Furthermore the sequence of the different loading types is of importance. Due to the prior loading with hydrogen plasma the thermal shock cracks were not limited to the electron beam loaded area but propagated through the just plasma loaded area and even through the unexposed area at the edges of the tungsten target.     

Simulation study of carbon impurity dynamics on tungsten surfaces exposed to hydrogen ions

The depth profile of C impurity deposited on a W target exposed to H⁺ and C⁺ impurities at a concentration of C: 0.8% has been calculated in terms of segregation, diffusion and chemical erosion. For the segregation, the Gibbsian model has been used. For the diffusion, a concentration dependent diffusion model (C in WC and/or C) has been utilized. For the chemical erosion, the chemical erosion yield much lower than that for the H-C system has been applied. The calculated depth profiles at 653 K and 913 K are in good agreement with the XPS data. The agreement indicates that there is a significant contribution of segregation, which shifts the maximum C concentration to the top surface in the depth profiles. On the other hand, there are little contributions from diffusion and chemical erosion, which are related closely to formation of WC in the target.     

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.     

Changed structure and improved operation characteristics of metals and alloys exposed to high power ion beams

The effect of high power ions beams (HPIB) on the structure of Al and Cu metals and P6M5 steel (8.5 × 10⁻⁴% C, 5% Mo, 6% W, 4% Cr, 2% V, 72% Fe) is presented. The thickness of the modified layer in Al and Cu was 150 to 180 μm. Defects observed in the bulk of the sample were shown to increase the microhardness and abrasive wear resistance of the samples. These defects are dislocation loops and prisms of vacancy or may be of interstitial type. The current density (or total deposited energy) was found to be the parameter most strongly influencing the target modification. To investigate the near surface layers, we used positron annihilation and Auger spectroscopy. Further, we measured microhardness and abrasive wear resistance.     

Conceptual design and safety characteristics of a new multi-mission high flux research reactor

Research reactors with neutron fluxes higher than 10¹⁴ n cm^-2 s^-1 are widely used in nuclear fuel and material irradiation,neutron-based scientific research, and medical and industrial isotope production. Such high flux research reactors are not only important scientific research facilities for the development of nuclear energy but also represent the national comprehensive technical capability. China has several high flux research reactors that do not satisfy the...     

Characteristics of GaAs MESFET inverters exposed to high energyneutrons

GaAs MESFET circuits were exposed to high-energy neutrons with fluences ranging from 1×10¹⁴ n/cm² to 2×10¹⁵ n/cm². The reflections of discrete transistors, inverters, and ring oscillators were characterized at each fluence. While the MESFETs exhibit significant threshold voltage shifts and transconductance and saturation current degradation over this range of neutron fluences, an improvement in the DC characteristics of Schottky diode FET logic (SDFL) inverters was observed. This unusual result has been successfully simulated using device parameters extracted from FETs damaged by exposure to high-energy neutrons. Although the decrease in device transconductance results in an increase in inverter gate delay as reflected in ring oscillator frequency measurements, it is concluded that GaAs ICs fabricated from this logic family will remain functional after exposure to extreme neutron fluences. This is a consequence of the observed improvement in inverter noise margin evident in both measured and simulated circuit performance     

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.     

On the multiaxial behaviour of concrete exposed to high temperature

An attempt is made to formulate a multiaxial constitutive model for concrete in the temperature range up to 800°C. The proposed model can be characterized as isotropic, elastic-viscoplastic-plastic in the compression region. Brittle failure is assumed in the tensile region.The thermal strain increment is assumed to be a function of both temperature and the current stress tensor. This assumption implies that the thermal strain may have deviatoric components.The volumetric thermal strain is used as a scalar damage measure instead of temperature itself. The corresponding softening function is obtained from isothermal, uniaxial tests. Also the elastic properties are taken as functions of the volumetric thermal strain.The response of the model is illustrated and compared with experimental results.