Hydrogen and helium trapping in tungsten under single and sequential irradiations

The trapping of hydrogen and helium in polycrystalline tungsten irradiated with 500 eV He⁺, H⁺ and D⁺ ions, individually or sequentially, has been measured by thermal desorption spectroscopy. Specimens irradiated with 500 eV He⁺ at 300 K show three He release peaks in the vicinity of ∼500, ∼1000, and ∼1200 K. The helium is thought to form He vacancy complexes or bubbles. Increasing the specimen temperature to 700 K does not significantly affect the trapping behavior of He. Sequential He⁺-D⁺ irradiation at 300 K results in the elimination of He release above 800 K. Instead, both D and He were released in the range 400-800 K. This is interpreted as interstitial D and He released from the near surface. Sequential He⁺-D⁺ irradiation at 700 K resulted in a reduced single He peak at ∼1000 K with very little release observed below 800 K; no D was trapped for irradiations at 700 K. Sequential D⁺-He⁺ irradiations at 300 K show that He trapping occurs in much the same manner as for the He⁺-only case while D retention is reduced at the near surface. Sequential D⁺-He⁺ irradiations at 700 K indicate that pre-irradiation with D⁺ has little or no effect on the subsequent trapping behavior of He.     

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.     

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.     

In situ ERDA measurements of hydrogen isotope concentrations in palladium at atmospheric pressure

In situ elastic recoil detection analysis (ERDA) measurements in gases at atmospheric pressure have been carried out using 15 MeV ⁴He ion beams. The beams are extracted through a molybdenum foil having a thickness of 5 μm. The maximum depth of analysis is about 4 μm for the palladium hydride and palladium deuteride (PdH_x and PdD_x, x = 0.7-0.8) samples. The temperature of the samples rises stepwise from room temperature to 180 °C. ERDA spectra are obtained every 2 min. Hydrogen and deuterium in the samples are discharged in the temperature range of 120-140 °C in a vacuum. Decrease in the hydrogen concentration in the PdH_x sample heated in a vacuum follows a first order in the value of x and an apparent activation energy of discharge of hydrogen is 1.05 eV. On the other hand, the hydrogen and deuterium concentrations decrease at about 80 °C in air. No isotope effects are observed in both a vacuum and air. The temperature at which the hydrogen concentration decreases in helium gas is almost the same as that in a vacuum. It indicates that hydrogen and deuterium atoms are discharged by chemical reactions with air and that there are no effects of cooling of the thermocouple by convection of air.     

Modelling deuterium release during thermal desorption of D-irradiated tungsten

Thermal desorption profiles were modelled based on SIMS measurements of implantation profiles and using the multi-trap diffusion code TMAP7 [G.R. Longhurst, TMAP7: Tritium Migration Analysis Program, User Manual, Idaho National Laboratory, INEEL/EXT-04-02352 (2004)]. The thermal desorption profiles were the result of 500 eV/D⁺ irradiations on single crystal tungsten at 300 and 500 K to fluences of 10²²-10²⁴ D⁺/m². SIMS depth profiling was performed after irradiation to obtain the distribution of trapped D within the top 60 nm of the surface. Thermal desorption spectroscopy (TDS) was performed subsequently to obtain desorption profiles and to extract the total trapped D inventory. The SIMS profiles were calibrated to give D concentrations. To account for the total trapped D inventory measured by TDS, SIMS depth distributions were used in the near-surface (surface to 30 nm), NRA measurements [V.Kh. Alimov, J. Roth, M. Mayer, J. Nucl. Mater. 337-339 (2005) 619] were used in the range 1-7 μm, and a linear drop in the D distribution was assumed in the intermediate sub-surface region (∼30 nm to 1 μm). Traps were assumed to be saturated so that the D distribution also represented the trap distribution. Three trap energies, 1.07 ± 0.03, 1.34 ± 0.03 and 2.1 ± 0.05 eV were required to model the 520, 640 and 900 K desorption peaks, respectively. The 1.34 and 1.07 eV traps correspond to trapping of a first and second D atom at a vacancy, respectively, while the 2.1 eV trap corresponds to atomic D trapping at a void. A fourth trap energy of 0.65 eV was used to fit the 400 K desorption peak observed by Quastel et al. [A.D. Quastel, J.W. Davis, A.A. Haasz, R.G. Macaulay-Newcombe, J. Nucl. Mater. 359 (2006) 8].     

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.     

Incorporation of transparent conducting oxide characteristics and oxygen analysis in vacuum annealed indium oxide films

Indium oxide films prepared by thermal oxidation of In metal films acquire high electrical conductivity and optical transmission on annealing in vacuum. These films, by virtue of their high figure of merit (1.3 × 10³ Ω⁻¹ cm⁻¹), can function as transparent conducting oxides. Deposition of smooth In metal films is essential for realizing good figure of merit in oxide films. Roughness existing in metallic films persists in oxide films and impairs their properties, in particular, optical transmission. Analyses by Rutherford backscattering spectrometry and ¹⁶O(α,α)¹⁶O resonant scattering suggest that vacuum annealing brings about a subtle oxygen deficiency in films and also causes a reduction in their thicknesses. The deficiency of oxygen results from the loss of lattice oxygen and desorption of chemisorbed oxygen, and is responsible for the enhanced electrical conductivity of the films. The reduction in thickness, on the other hand, occurs due to the volatilization of In₂O formed under the reducing environment of vacuum. X-ray photoelectron spectroscopy studies show that vacuum annealing renders films reactive towards oxygen bearing species. These films, as a result, tend to regain the nominal oxide composition on prolonged storage in ambient air, however, their TCO properties remain largely unaffected. The nature of oxygen absorbed in the process is seemingly different from that in air-annealed oxide films.     

Investigations of adsorption states of protium and deuterium in redeposited carbon flakes formed in tokamak T-10

This work reports thermal desorption spectroscopy (TDS), Fourier-transform infrared (FT-IR) spectroscopy and X-ray diffraction (XRD) analyses on free standing redeposited hydrocarbon films (flakes) with a high deuterium to hydrogen isotopic ratio, produced in the T-10 tokamak in the Kurchatov Institute. XRD pattern showed that the carbon flakes differ substantially from graphite and are non-crystalline. The TDS D₂(H₂) curves consist of two groups of peaks (450-800 K and 900-1000 K), and appeared to be rather similar to those obtained for a mechanically milled nanostructured graphite. As a result, two main adsorption states with activation energies of about 0.65 and 1.25 eV/H were found, implying a hopping diffusion and a resonance mechanism, respectively. The IR spectral differences between reddish-gold and dark-brown flakes showed a less degree of C-H hybridization for dark films and a disordered carbon network, to which the CD₂,₃ end-groups are connected.     

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.     

材料中氢同位素行为热脱附谱实验方法研究

材料中氢同位素行为研究是确保聚变堆安全和经济性的关键问题和重要研究方向。为研究材料中氢同位素的扩散、释放、居留等特性,建立了一种联合四极质谱仪（QMS）的热脱附谱（TDS）实验方法,解决了TDS系统超高真空、低氢同位素质谱本底、线性升温速率控制以及灵敏度标定等关键科学技术问题。通过涡轮分子泵和二级溅射离子泵实现了优于1×10^-7Pa的超高真空,本底H₂分压降至1×10^-9Pa。通过MCGS直流PID控温程序实现样品升温速率在1～100 K/min范围可调,采用漏率可变的特制通导型玻璃漏孔标定TDS系统的氘气脱附速率灵敏度,确定该灵敏度系数α和最小可检测氘气热脱附速率（脱附速率灵敏度）分别为6.22×10²⁴s^-1·A^-1、1.24×10^-10s^-1。采用镀镍Zr-4合金吸氘样品验证了TDS方法的有效性,初步分析了Zr-4中的氘热脱附特性。     

Fuel retention study in fusion reactor walls by micro-NRA deuterium mapping

Nuclear Reaction Analysis (NRA) with a ³He ion beam is a powerful analytical technique for analysis of light elements in thin films. The main motivation for ³He focused beam applications is lateral mapping of deuterium using the nuclear reaction D(³He,p)⁴He in surfaces exposed to a tokamak plasma, where a lateral resolution in the μm-range provides unique information for fuel retention studies.At the microprobe at the Jo?ef Stefan Institute typical helium ion currents of 300 pA and beam dimensions of 4 × 4 μm² can be obtained. This work is focused on micro-NRA studies of plasma-facing materials using a set-up consisting of a silicon partially depleted charge particle detector for NRA spectroscopy applied in parallel with a permanently installed X-ray detector, an RBS detector and a beam chopper for ion dose monitoring. A method for absolute deuterium quantification is described. In addition, plasma-deposited amorphous deuterated carbon thin films (a-C:D) with known D content were used as a reference.The method was used to study deuterium fuel retention in carbon fibre composite materials exposed to a deuterium plasma in the Tore Supra and TEXTOR tokamaks. The high lateral resolution of micro-NRA allowed us to make a detailed study of the influence of topography on the fuel retention process. We demonstrated that the surface topography plays a dominant role in the retention of deuterium. The deep surfaces inside the castellation gaps showed approximately two orders of magnitude lower deuterium concentrations than in areas close to the exposed surface.     

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.     

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).     

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 and Physical Sputtering of Low Activation Ferritic Steel

Low activation materials have to be developed toward fusion demonstration reactors. Ferritic steel, vanadium alloy and SiC/SiC composite are candidate materials of the first wall, vacuum vessel and blanket components, respectively. Although changes of mechanical-thermal properties owing to neutron irradiation have been investigated so far, there is little data for the plasma material interactions, such as fuel hydrogen retention and erosion. In the present study, deuterium retention and physical sputtering of low activation ferritic steel, F82H, were investigated by using deuterium ion irradiation apparatus. After a ferritic steel sample was irradiated by 1.7 keV D^ ions, the weight loss was measured to obtain the physical sputtering yield. The sputtering yield was 0.04, comparable to that of stainless steel. In order to obtain the retained amount of deuterium, technique of thermal desorption spectroscopy (TDS) was employed to the irradiated sample. The retained deuterium desorbed at temperature ranging from 450 K to 700 K, in the forms of DHO, D2, D2O and hydrocarbons. Hence, the deuterium retained can be reduced by baking with a relatively low temperature. The fiuence dependence of retained amount of deuterium was measured by changing the ion fiuence. In the ferritic steel without mechanical polish, the retained amount was large even when the fluence was low. In such a case, a large amount of deuterium was trapped in the surface oxide layer containing O and C. When the fluence was large, the thickness of surface oxide layer was reduced by the ion sputtering, and then the retained amount in the oxide layer decreased. In the case of a high fluence, the retained amount of deuterium became comparable to that of ferritic steel with mechanical polish or SS 316 L, and one order of magnitude smaller than that of graphite. When the ferritic steel is used, it is required to remove the surface oxide layer for reduction of fuel hydrogen retention. Ferritic steel sample was exposed to the environment of JFT-2M tokamak in JAERI and after that the deuterium retention was examined. The result was roughly the same as the case of deuterium ion irradiation experiment.     

D2 gas-filled blisters on deuterium-bombarded tungsten

Most of spherical blisters formed by deuterium (D) bombardment (38 eV/D) up to 3 × 10²⁴ D/m² at 300 K on polycrystalline tungsten are fully elastic deformations. This has been proven by opening individual blisters with a focused ion beam and in situ observation of their complete relaxation by scanning electron microscopy. The D₂ gas filling is confirmed by observing simultaneously the D₂ puff. The gas pressure is causal for the stability of such spherical blisters after implantation and the gas release leads to sudden relaxation. The dilatation of the blister cap by trapped D can be excluded as cause for the blisters.     

Oxygen glow discharge experiment to remove deposited layers and to release trapped hydrogen isotopes in HT-7 superconducting tokamak

An experiment to remove re-deposited layers and to release hydrogen using a glow discharge in oxygen (O-GDC) has been performed in the HT-7 superconducting tokamak. In the absence of magnetic fields, the O-GDC wall conditioning had produced rapid, controlled co-deposit removal. Average removal rates, 5.2 × 10²² H-atoms/h, 5.65 × 10²¹ D-atoms/h and 5.53 × 10²² C-atoms/h, respectively, were obtained during 145 min O-GDC experiment in the pressure range 0.5-1.5 Pa. The corresponding removal rate of co-deposited films was ∼1.19 μm/day (26.5 g/day for carbon) based on an area of 12 m². Compared to thermo-oxidation and O-ICR experiment, high pressure O-GDC wall conditioning promoted the oxidation and improved the C and D atoms removal. In the O-GDC experiment, the removal rates of H-atoms and D-atoms as H₂O, HDO and D₂O were higher than that of H₂ and D₂ by factors of about 20 and 50, respectively. During the 145 min O-GDC experiment, about 14.5% O-atoms were converted into carbon oxides and hydroxides, and about 5.37 × 10²² O-atoms were adsorbed on the walls corresponding to a coverage of 4.5 × 10²¹ O/m² on an wall area of 12 m². In a 100 min helium glow discharge (He-GDC) following the O-GDC experiment, 1.53 × 10²² O-atoms, about 28.5% oxygen retained on the walls, were removed. The removal rate of H-atoms in He-GDC cleaning after O-GDC experiment was lower than that in He-GDC cleaning before O-GDC experiment, which indicates that the O-GDC wall conditioning had effectively reduced hydrogen retention on the walls.     

Interaction of hydrogen plasma with carbon-tungsten composite layer

Interaction of neutral hydrogen atoms with the layer of hydrogenated carbon-tungsten composite was studied. A 1 μm thick layer was prepared by sputter deposition from C and WC-targets in Ar/C₂-H₂ gas mixture. After deposition the samples were treated in weakly ionized highly dissociated hydrogen plasma created in a microwave discharge at a power of 1 kW. The gas flow was 13 l/h and pressure was 90 Pa. Temperature of the samples during treatment was about 850 K. After plasma treatment the samples were analyzed by AES (Auger electron spectroscopy) depth profiling, XPS (X-ray photoelectron spectroscopy) and SEM (scanning electron microscopy). It was found that during hydrogen plasma treatment selective etching of the C-W layer occurred. Carbon was preferentially removed from the C-W layer, and after about 10 min of treatment practically only tungsten with a huge porosity was detected.     

Dynamic Monte-Carlo modeling of thermal desorption of H2 from H irradiated graphite

Thermal desorption of hydrogen molecules from H⁺ irradiated graphite is studied using dynamic Monte Carlo simulation. The purpose of this study is to understand the experimentally observed phenomena that the thermal desorption of H₂ from the graphite exhibits sometimes single desorption peak, sometimes double peaks, and even three desorption peaks under certain circumstances. The study result reveals that the fluence of pre-implanted H⁺, the concentration of trap sites, porosity, and mean crystallite volume are important parameters in determining the number of desorption peaks. It is found that low implantation fluence and high concentration of trap sites easily lead to the occurrence of single desorption peak at around 1000 K, and high implantation fluence and low concentration of trap sites favor the occurrence of double desorption peaks, with a new desorption peak at around 820 K. It is also found that small porosity of graphite and large crystallite volume benefit the occurrence of single desorption peak at around 1000 K while large porosity of graphite and small crystallite volume facilitate the occurrence of double desorption peaks, respectively, at around 820 and 1000 K. In addition, experimentally observed third desorption peak at lower temperature is reproduced by simulation with assuming the graphite containing a small concentration of solute hydrogen atoms.     

RBS study of Ti/ZnO interface

We have studied the interface stability of the Ti(overlayer)/ZnO(substrate) system. Ti thin film was grown on the Zn face of single crystal ZnO(0 0 0 1) substrate by the vacuum deposition technique. The Ti film thickness was typically 16 nm. Then the samples were annealed in air at 300 and 400 °C for 15 min, respectively. The deposition and annealing effects on the interface structure were investigated with Rutherford backscattering and channeling spectroscopy using 2 MeV He⁺ ion beam. After Ti deposition the minimum yield from the ZnO substrate increased from 2% to 7%. This suggests severe damage caused by deposition, i.e. the interface reaction between Ti and ZnO (even at room temperature). A significant amount of Zn (approximately 6.4 × 10¹⁶ atoms/cm²) moved onto the surface after post-annealing at 400 °C. Since Ti has a stronger tendency to react with O than Zn, it is expected that Ti reacts with substrate oxygen leaving behind free Zn atoms, which can easily migrate onto the surface. We discuss how the Ti/ZnO interface reaction in detail, and seek to find another good metallic contact for ZnO devices, which are attracting much attention recently for practical applications as well as scientific aspects.