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.     

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.     

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.     

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.     

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.     

Deuterium retention in sintered boron carbide exposed to a deuterium plasma

Depth profiles of deuterium up to a depth of 10 μm have been measured using the D(³He,p)⁴He nuclear reaction in a resonance-like technique after exposure of sintered boron carbide, B₄C, at elevated temperatures to a low energy (≈200 eV/D) and high ion flux (≈10²¹ m⁻² s⁻¹) D plasma. The proton yield was measured as a function of incident ³He energy and the D depth profile was obtained by deconvolution of the measured proton yields using the program SIMNRA. D atoms diffuse into the bulk at temperatures above 553 K, and accumulate up to a maximum concentration of about 0.2 at.%. At high fluences (?10²⁴ D/m²), the accumulation in the bulk plays a major role in the D retention. With increasing exposure temperature, the amount of D retained in B₄C increases and exceeds a value of 2 × 10²¹ D/m² at 923 K. The deuterium diffusivity in the sintered boron carbide is estimated to be D = 2.6 × 10⁻⁶exp{−(107 ± 10) kJ mol⁻¹/RT} m² s⁻¹.     

Surface modifications and deuterium depth profiles in molybdenum irradiated with low-energy D ions

Depth profiles of deuterium trapped in single crystal Mo, polycrystalline Mo, and molybdenum trioxide film on polycrystalline Mo irradiated with 200 eV D ions have been measured up to a depth of 8 μm using the D(³He,p)⁴He nuclear reaction at a ³He energy varied from 0.69 to 4.0 MeV. For the D ion irradiation at 323 K to the highest ion fluence of 5 × 10²⁴ D/m², the D concentration decreases from several at.% in the near-surface layer to bulk values below 10⁻⁴ at.% for single crystal Mo and about 10⁻² at.% for polycrystalline Mo. The maximum D concentration in molybdenum trioxide film differs little in value from that for polycrystalline Mo. Blister formation at high fluences is observed for polycrystalline Mo and molybdenum trioxide film, but not for single crystal Mo. As the irradiation temperature increases from 323 to 493 K, the D retention in the polycrystalline Mo decreases from about 3 × 10²¹ down to about 2 × 10¹⁸ D/m².     

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.     

Deuterium retention and release in tungsten co-deposited layers

A systematic study of the influence of the deposition conditions on the deuterium retention in co-deposited tungsten layers formed both by magnetron sputtering and in the PISCES-B linear device has been carried out. Experimental parameters such as the tungsten deposition rate, the incident particle energy and the substrate temperature are shown to affect the level of deuterium retention in the layers. A decreased retention for increased substrate temperature and deposition rates, and an increased retention for increasing incident deuterium particle energy are observed. A scaling equation is proposed to describe the influence of the conditions during the co-deposition process (surface temperature, incident particle energy and deposition flux) on the deuterium retention. In addition, the desorption kinetics of deuterium has been studied by TDS. Two desorption stages at 473-573 K and at 1073 K have been observed.     

Surface morphology influence on deuterium retention in beryllium films prepared by thermionic vacuum arc method

In a plasma-confinement device, material eroded from plasma facing components will be transported and re-deposited at other locations inside the reaction chamber. Since beryllium from the first wall of the ITER fusion reactor will be eroded, ionized in the scrape-off layer plasma and finally re-deposited on divertor surfaces flowing along the magnetic field, it is important to study the properties of divertor armour materials (C, W) coated with beryllium.By applying different bias voltages (−200 V to +700 V) to the substrates during deposition, the morphology of the obtained films was modified. The films’ morphology was characterized by means of AFM and SEM, and it was found that the coatings prepared using negative bias voltage at the substrate during deposition are more compact and have a smoother surface compared to the samples prepared with positive bias voltage. The thickness and composition of each film were measured using Rutherford backscattering spectrometry (RBS). A study of deuterium implantation and retention into the prepared films was performed at IPP Garching in the high current ion source.     

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.     

Influence of temperature and plasma composition on deuterium retention in refractory metals

Refractory materials are being considered potential candidates to build the first wall of the fusion reactor chamber. This work reports on the results of the study of tungsten and molybdenum metals exposed to high flux densities (～10²⁴ D/m² s) and low temperature (T_e ～ 3 eV) deuterium plasmas in Pilot-PSI irradiation facility.The hydrogenic retention in poly-crystalline W and Mo targets was studied with ³He nuclear reaction analyses (NRA). The NRA results clearly show a two-dimensional radial distribution of the deuterium with a minimum at the center and a maximum close to the edge. These distribution correlates well with the thermal profile of the sample surface, where a maximum of ～1600 K was measured at the center decreasing to ～1000 K in the edges. A maximum deuterium fluence retention of 5 × 10¹⁵ D/cm² was measured. The values of the retained fractions ranging from 10^?5 to 10^?6 D_retained/D_incident were measured with thermal desorption spectroscopy (TDS) and compares well with IBA results. Moreover, the presence of C in the plasma and its co-deposition increases the D retention in the region where a C film is formed. Both NRA and TDS results show no clear dependence of retention on incident fluence suggesting the absence of plasma related traps in W under these conditions.     

Deuterium retention in tungsten at fluences of up to 10 D/m using D ion beams

Deuterium retention in two types of polycrystalline tungsten (PCW) was studied as a function of incident ion fluence, ion energy, and specimen temperature. (i) D retention at 300 K, as a function of D⁺ fluence, demonstrated a trend to saturation in both the Rembar hot-rolled thin foil and Plansee tungsten plate. At 500 K, new D retention results for the Plansee PCW showed an increasing trend with increasing incident D⁺ fluence without any indication of saturation, in agreement with previous results for Rembar PCW [A.A. Haasz, J.W. Davis, M. Poon, R.G. Macaulay-Newcombe, J. Nucl. Mater. 258-263 (1998) 889-895]. Even when the incident D⁺ fluence was increased to 8 × 10²⁵ D⁺/m², which is in the fluence range of plasma devices, there was still no sign of saturation. (ii) The temperature dependence results for the Plansee PCW show a decreasing trend in D retention as the temperature is increased from 300 to 500 K. These results differ from previous studies of Rembar PCW [A.A. Haasz, J.W. Davis, M. Poon, R.G. Macaulay-Newcombe, J. Nucl. Mater. 258-263 (1998) 889-895], but are similar to those seen for single crystal tungsten [M. Poon, A.A. Haasz, J.W. Davis, R.G. Macaulay-Newcombe, J. Nucl. Mater. 313-316 (2003) 199]; an explanation for the different behaviour is suggested. (iii) Varying the D⁺ energy from 100 to 500 eV/D⁺ plays a minor role in the amount of D retained, suggesting that D retention in W depends more on the W structure, incident ion fluence and specimen temperature, rather than on the incident ion energy when the energy is below the threshold for damage formation (∼960 eV for D on W).     

Localization by TEM and EELS of deuterium trapping sites in CFC exposed to plasma irradiation in Tore Supra

Carbon fiber composite (CFC) Sepcarb^® N11 is used in the tokamak Tore Supra as plasma-facing components. To investigate the fuel retention capability of this material, a mobile sample holder was used to expose CFC N11 samples to direct irradiation by the scrape-off layer plasma of Tore Supra at fluences up to 1 × 10²⁵ m⁻². Deuterium (D) elemental mapping using nuclear reaction analysis for the most-exposed CFC sample showed that D retention occurs at depths greater than 8 μm due to the presence of deep (>3.5 μm) local retention sites. In this work, combining transmission electron microscopy (TEM) and electron energy-loss spectroscopy (EELS), we describe at a high spatial resolution where and how D atoms are trapped in these sites. TEM experiments performed on thin cross-sections of the plasma-modified surface show evidence of the presence of a 3.5 μm-thick deuterated amorphous carbon layer deposited on the CFC surface. We show that specific localized retention sites correspond to the filling of relatively large (∼3 μm.) and deep (at least 3 μm below the initial CFC surface) cracks between fibres and matrix by the deuterated amorphous carbon layer.     

Study on the retention behavior of hydrogen isotopes and the change of chemical states of boron film exposed to hydrogen plasma in LHD

The behavior of hydrogen retention and the change of chemical states of boron film exposed to hydrogen plasma in LHD were investigated. The sample was prepared in LHD, and atomic concentrations for the boron film after hydrogen plasma exposure were changed from 75% for boron, 15% for carbon and 8% for oxygen to 53%, 18% and 22%, respectively. BC bond was a major chemical state of the boron film after hydrogen plasma exposure, although abundance of BB bond was the highest before the plasma exposure. Total hydrogen retention measured by TDS was evaluated to be 1.7 × 10²⁰ H m^?2, and the retentions of hydrogen as BHB, BH and BCH bonds were, respectively, 4.8 × 10¹⁹, 7.2 × 10¹⁹ and 5.2 × 10¹⁹ H m^?2. It was concluded that the hydrogen retention could be estimated by taking account not only of chemical states of impurities, but also of hydrogen depth profile.     

Tritium uptake in graphite tiles exposed to EAST plasma and then tritium gas

Tritium exposure experiments were carried out for three kinds of EAST SiC coated doped-graphite (SiC/C) samples, one from the original graphite tiles without being irradiated, and the other two from erosion and deposition areas of first wall after the 2009 campaign in EAST. β-ray-induced X-ray spectrometry (BIXS) was used to characterize the exposed samples. It is showed that the significant amount of tritium was absorbed in the surface of deposition sample in comparison with that of original sample, which was also supported by the results of imaging plate (IP) measurements. In addition, it was found that drastic decrease in tritium retention appeared by lowering exposure temperature, and the trapped tritium was maintained stably with time. Computer simulation is used to analyze the details of depth profile of tritium in different kinds of samples.     

Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

The influence of the oxide layer morphology on the hydrogen uptake during steam oxidation of (Zr,Sn) and Zr-Nb nuclear fuel rod cladding alloys was investigated in isothermal separate-effect tests and large-scale fuel rod bundle simulation experiments. From both it can be concluded that the concentration of hydrogen in the remaining metal strongly depends on the existence of tangential cracks in the oxide layers formed by the tetragonal - monoclinic phase transition in the oxide, known as breakaway effect. In these cracks hydrogen is strongly enriched. It results in very local high hydrogen partial pressure at the oxide/metal interface and in an increase of the hydrogen concentration in the metal at local regions where such cracks in the oxide layer exist. Due to this effect the hydrogen uptake of the remaining zirconium alloy does not depend monotonically on temperature. Differences between (Zr,Sn) and Zr-Nb alloys are caused by differences in the hydrogen production due to different oxidation kinetics and in the crack forming phase transformation in the oxides as well as in the mechanical stability of the oxides.     

Grazing-incidence electron microscopy of surface blisters in single- and polycrystalline tungsten formed by H, D and He irradiation

Blisters on single- and polycrystalline tungsten surfaces formed by hydrogen and helium ion irradiation were investigated by grazing-incidence electron microscopy (GIEM) with an ultra-high-voltage transmission electron microscope. It was found that the blister skin thickness formed by D⁺ irradiation of polycrystalline tungsten (PCW) was considerably larger than the calculated ion range of the implants; however, this skin thickness (or blister depth) is not related to the pre-existing grain boundaries in the PCW. Blister formation was also observed with GIEM for single crystal tungsten (SCW) irradiated with H⁺, D⁺, and He⁺. The critical ion fluence for blister formation in SCW is estimated to be ∼10²³ H⁺(D⁺)/m² for H(D) and ∼10²¹ He⁺/m² for He. The size of the blisters and their skin structure depends on the irradiating conditions. Typical skin thickness was about 50-150 nm. Based on the assumption that gas particles (H₂, D₂, and He) accumulate within the blisters during H⁺, D⁺, and He⁺ irradiation, the GIEM measurements provide a means to derive an estimate of the amount of gas so accumulated, by reproducing the observed blister shapes with finite element method (FEM) calculations. From the GIEM images and FEM calculations we have estimated the number of implanted ions being retained in the blisters, and compared these amounts with published retention measurements. A mechanism for the blister formation is proposed based on the present results.