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We examine the effect of neutron irradiation on the release of deuterium from tungsten at 573 K to understand the efficiency of tritium removal by baking out at moderate temperatures. Tungsten samples, undamaged and neutron-irradiated to a damage level of approximately 0.016 displacements per atom, are exposed to low-energy (108 eV), high-flux (3.0 × 1021 to 9.4 × 1021 m−2 s−1) deuterium plasma at temperatures ranging from 573 to 773 K to an ion fluence of 1.1 × 1025 m−2. At each exposure temperature, two undamaged and two neutron-irradiated tungsten samples are exposed to plasma. The deuterium content in the tungsten samples is measured by thermal desorption spectrometry soon after the plasma exposure and after post-plasma annealing at 573 K for 30 h. It is found that: (i) the deuterium retention in the neutron-irradiated tungsten samples is significantly higher than that in the undamaged tungsten samples; (ii) annealing at 573 K of undamaged tungsten samples pre-exposed to deuterium plasma at 573–773 K leads to an almost complete (60%–99%) release of deuterium from the samples; (iii) annealing at 573 K of neutron-irradiated tungsten samples pre-exposed to deuterium plasma at 573–773 K leads to a significant (8%–20%) release of deuterium from the samples. [ABSTRACT FROM AUTHOR]

Eight samples of F82H ferritic/martensitic steel were irradiated at 300 K with 20 MeV W ions to the damage level of 0.54 displacements per atom at the damage peak. Three samples were afterwards annealed in vacuum at 423 K for 72 h and then at 373 K for 106 h. Three other samples were annealed in H 2 atmosphere at 100 kPa at the same annealing temperatures and durations. All samples were exposed at room temperature to D 2 O vapor at the partial pressures in the range from 3 to 6 kPa for 365 to 1181 days. After termination of the D 2 O vapor exposure, the surfaces of the samples were partly covered by small (≤1 mm in diameter) drops of D 2 O water. Thus, the damaged samples were exposed to a mixture of D 2 O vapor and water. Trapping of deuterium at the ion-induced defects in the damage zone was examined by the D( 3 He, p) 4 He nuclear reaction. It has been found that the W-ion-induced defects generated in the F82H samples are decorated by deuterium diffusing from the surface. [ABSTRACT FROM AUTHOR]

Surface topography and deuterium retention in polycrystalline hot-rolled W and W–5%Re have been examined after exposure to a low-energy (76 eV), high flux (around 10 22 D/m 2 s) deuterium plasma to an ion fluence of 10 26 D/m 2 at various temperatures. The methods used were confocal laser scanning microscopy and the D( 3 He, p) 4 He nuclear reaction at 3 He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 348 to 673 K, small blisters of size in the range from about 1 to about 15 μm, depending on the exposure temperature, are formed on the W and W–5%Re surfaces. In the W–5%Re, the deuterium retention demonstrates its maximum at exposure temperature of 463 K, while in the W this maximum is shifted to 523 K. A difference in the temperature dependence of the D retention for the W and W–5%Re is explained, as a rough approximation, by temperature dependences of the ductility of these materials. [ABSTRACT FROM AUTHOR]

Modelling tritium retention in radiation-damaged pure-tungsten samples, the concentration of deuterium retained in the tungsten-ion-induced damage zone decreases with increasing exposure temperature; whereas for a similarly treated tungsten–rhenium alloy, it can drop significantly faster than in pure tungsten, given sufficiently high temperatures. In contrast, similar changes in retention behaviour were not observed in undamaged samples. Based on these findings, as well as on corresponding previous TEM results, it is concluded that the concentration of high-energy radiation-induced defects responsible for trapping of deuterium is lower in the alloy than in pure tungsten. Therefore, a small amount of transmutation rhenium in damaged tungsten should be able to keep tritium retention to a low level in ITER. [ABSTRACT FROM AUTHOR]

Highlights: [•] W samples were pre-irradiated with 4.8 and 20MeV W ions to various displacements per atom (dpa). [•] Under following exposure to D plasmas radiation-induced defects were occupied by diffusing D atoms. [•] At ⩾0.1dpa, the D concentration in the damage zone demonstrated weak dependence on the damage level. [ABSTRACT FROM AUTHOR]

Abstract: Surface topography and deuterium retention in polycrystalline ITER-grade tungsten have been examined after exposure to a low-energy (38eV/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 4.0MeV. During exposure to the D plasma at temperatures in the range from 320 to 815K, small blisters of size in the range from 0.2 to 5μm, 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 500K, and then decreases below 1019 D/m2 at 800K. [Copyright &y& Elsevier]

Abstract: Samples prepared from polycrystalline ITER-grade tungsten were damaged by irradiation with 20MeVW 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, 1.35μm 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 4.0MeV allowing determination of the D concentration at depths up to 6μm. It has been found that (i) addition of 10% helium ions into the D plasma at exposure temperatures of 440–650K significantly reduces the D concentration at depths of 0.5–6μm 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 2μm (i.e., in the damage zone) under subsequent exposures to both pure D and D–He plasmas. [Copyright &y& Elsevier]

Abstract: Surface topography and deuterium retention in tungsten have been examined after exposure to a low-energy (38eV/D), high-flux (1022 D/m2s) deuterium plasma at ion fluences of 1026 and 1027 D/m2 and 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 4.0MeV. During exposure to the D plasma, blisters with various shapes and sizes depending on the exposure temperature are formed on the W surface. At the temperatures above 700K no blisters appear. The deuterium retention increases with the exposure temperature, reaching its maximum value of about 1022 D/m2 at 480–530K, and then decreases to about 1019 D/m2 at 800K. [Copyright &y& Elsevier]

Abstract: Two types of porous plasma spray tungsten coatings deposited onto stainless steel and graphite substrates were exposed to low-energy (76eV ), 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 7μm. The relatively high D concentration (above 0.1at.%) at depths of several micrometers observed after D plasma exposure at 340–560K can be related to accumulation of D2 molecules in pores, while at temperatures above 600K deuterium is accumulated mainly in the form of D atoms chemisorbed on the inner pore surfaces. At exposure temperatures above 500K, 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. [Copyright &y& Elsevier]

Abstract: The surface topography and optical properties of recrystallized tungsten exposed to a low-energy (38eV/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 320–695K with the exception of the narrow temperature region around 535K, 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 220–650nm, whereas the estimations of reflectance using the ellipsometry data demonstrate some increase. [Copyright &y& Elsevier]

Abstract: A study of the influence of the deposition conditions on the surface morphology and deuterium (D) concentration in tungsten (W) deposition layers formed by magnetron sputtering and in the linear plasma generator has been carried out. Thick W layers (⩾0.4μm) deposited onto copper substrates demonstrate areas of pilling and, after post-deposition heating to 1300K, flaking-off and fracturing. For thin W layers (⩽80nm) deposited onto stainless steel (SS) and W substrates, no areas of flaking-off and fracturing exist both after deposition and after post-deposition heating to 673K for the SS substrate and to 1300K for the W substrate. The concentration of deuterium in the W layers was found to decrease with increasing substrate temperature and with increasing tungsten deposition rate. For layers with relatively high concentration of oxygen (0.20–0.60O/W), a decrease of the D concentration with increasing substrate temperature is more pronounced than that for layers deposited in good vacuum conditions. To describe the evolution of the D/W ratio with the substrate temperature and the tungsten deposition rate, an empirical equation proposed by De Temmerman and Doerner [J. Nucl. Mater. 389 (2009) 479] but with alternative parameters has been used. [Copyright &y& Elsevier]

Abstract: Surface and sub-surface morphology and deuterium retention in polycrystalline tungsten, undamaged and pre-damaged with 5.5MeV W2+ ions to damage levels of 0.04, 0.4, and 1.2dpa, with and without subsequent annealing at 1200K for 2h, have been examined after irradiation with low-energy deuterium ions up to a fluence of 1×1026 D/m2 at various temperatures. The morphology changes were investigated by scanning electron microscopy in combination with focused ion beam surface cutting. The different hydrogen binding states were investigated by thermal desorption spectroscopy, and the depth profile of the implanted D was measured by ion beam nuclear reaction analysis. The dependence of the D retention on the damage level and the implanted fluence was studied, and the results show saturation for 0.4dpa independent of the deuterium implantation fluence. Additionally, it was observed that self-implantation introduces high temperature traps which can almost completely be removed by annealing at 1200K. [Copyright &y& Elsevier]

Abstract: First wall protection tiles manufactured from recrystallized graphite doped by Ti (RGTi) were extracted from the first wall of the Globus-M spherical tokamak after being used as plasma-facing components (PFCs) during more than 8000 pulses (more than 800s). The analysis of chemical composition and structure of the irradiated tiles was performed by different methods. It was found that most of the tiles were covered with mixed layers. In the deposits the amount of absorbed deuterium was measured and minimum D concentration was found in the region of maximum power loads. Only a small amount of deuterium has been detected in the bulk of RGTi tiles exposed to plasma. [Copyright &y& Elsevier]

Abstract: Depth profiles of deuterium trapped in single crystal Mo, polycrystalline Mo, and molybdenum trioxide film on polycrystalline Mo irradiated with 200eV D ions have been measured up to a depth of 8μm using the D(3He,p)4He nuclear reaction at a 3He energy varied from 0.69 to 4.0MeV. For the D ion irradiation at 323K 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 493K, the D retention in the polycrystalline Mo decreases from about 3×1021 down to about 2×1018 D/m2. [Copyright &y& Elsevier]

Abstract: Depth profiles of deuterium trapped in tungsten exposed to a low-energy (≈200eV/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 7μm using the D(3He,p)4He nuclear reaction at a 3He energy varied from 0.69 to 4.0MeV. The deuterium retention in single-crystalline and polycrystalline W increases with the exposure temperature, reaching its maximum value at about 500K (for ‘clean’ plasma) or about 600K (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. [Copyright &y& Elsevier]

Abstract: Plasma-facing carbon fibre composites (CFC) are used for Tore Supra (CFC N11), ASDEX Upgrade (CFC DMS 701) and foreseen for ITER (CFC NB31). Therefore, deuterium retention in the CFC NB31 and N11 was studied by irradiation with 30–200eV D ions, both from ion beams and plasma devices (PISCES-A, Magnetron plasma device), using methods of the thermal desorption spectrometry and D(3He,p)4He nuclear reaction in a resonance-like technique. It is found that at each irradiation temperatures in the range from 50 to 450°C the amount of deuterium trapped in the CFCs increases with the ion fluence, Γ. No saturation is reached as observed in pyrolytic graphite. At room temperature the retention at high fluence increases proportional to Γ 0.5. Depth profiles show that saturation occurs only within a near-surface layer corresponding to the ion range. The increase in total retention is accompanied by an increasing long profile tail extending beyond 14μm with the D concentration above 10−1 at.% at fluences above 1024 D/m2. Analysis of CFC and fine-grain graphite tiles removed from identical strike-point positions in the outer ASDEX Upgrade divertor show an order of magnitude higher retention in the CFC and profiles extending into the bulk with concentrations of 0.1at.% well beyond the measurable depth of 14μm. [Copyright &y& Elsevier]

Abstract: Depth profiles of deuterium up to a depth of 10μm 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 (≈200eV/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 553K, and accumulate up to a maximum concentration of about 0.2at.%. 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 923K. The deuterium diffusivity in the sintered boron carbide is estimated to be D =2.6×10−6exp{−(107±10)kJmol−1/RT}m2 s−1. [Copyright &y& Elsevier]

Abstract: The differential D(3He,p)4He reaction cross-section was determined at a laboratory angle of 135° for 3He energies between 0.5 and 6MeV. Deuterium depth profiles are determined from the measurement of the proton yield as a function of incident 3He energy with following deconvolution of the obtained yield function. This allows deuterium depth profiling to very large depths. Deuterium depth profiling to a depth of 7μm in W and 50μm in B4C is demonstrated. The accuracy of the method is discussed. [Copyright &y& Elsevier]

Abstract: Depth profiles of deuterium trapped in W single crystals and polycrystalline W irradiated with 200eV D ions at 300–323K have been measured up to a depth of 7μm using the D(3He,p)4He nuclear reaction in a resonance-like technique. The proton yield as a function of incident 3He energy was measured and the D depth profile was obtained by deconvolution of the measured proton yields using the program simnra. The depth at which deuterium is retained can be tentatively divided into three zones: (i) the near-surface layer (up to a depth of ∼0.2μm), (ii) the sub-surface layer (from ∼0.5 to ∼2μm), and (iii) the bulk (>5μm). Low-energy D ion irradiation modifies the W structure to depths of up to about 5μm, both for W single crystals and polycrystalline W. Up to this depth trapping sites are created at a fluence of 5×1022/m2 for the retention of deuterium at about 0.1at.%. [Copyright &y& Elsevier]

Hydrogen isotopes exchange in tungsten was investigated after sequential exposures to low energy deuterium (D) and helium–seeded protium (He–seeded H) plasmas at sample temperatures of 403 and 533 K. Deuterium depth profiles were measured by the D( 3 He, p) 4 He nuclear reaction with 3 He + energies between 0.69 and 4.5 MeV allowing determination of the D concentration up to a depth of 8 μm. It was found that a significant part of the deuterium initially retained in tungsten after D plasma exposure was released during sequential exposure to a protium plasma. However, exposure of the D-plasma-exposed W samples to the He–seeded H plasma reduces the amount of released deuterium as compared to pure H plasma exposure. [ABSTRACT FROM AUTHOR]

• W samples were irradiated with 20 MeV W ions to a displacement-damage level of 0.23 dpa. • The damaged W samples were exposed to a D 2 ― DT ― T 2 gas mixture at elevated temperatures. • Chemical etching was used to remove surface layers from the damaged side of these W samples. • Analysis of T in the etching solution made it possible to determine the T concentration in the etched layer. W samples, previously irradiated with 20 MeV W ions at room temperature to a displacement-damage level of 0.23 displacements per atom (dpa) at the peak of displacements, were exposed to a deuterium-tritium (D 2 ― DT ― T 2) gas mixture with a tritium content of 6% at a total pressure of 1.2 kPa and temperatures of 773 K and 973 K for 3 h. The concentration of tritium retained in the displacement-damage zone of these W samples was determined by a method combining chemical etching and subsequent analysis of tritium in the etching solution using liquid scintillation counting (CE-LSC). (Some figures may appear in colour only in the online journal) [ABSTRACT FROM AUTHOR]

• Ten Si probes were located on the outer side of the LHD first-wall surface in each 36° toroidal section. • During the 16th plasma campaign, mixed-material layers were deposited on the probes. • The mixed-material layers were examined by RBS/ERDA and cross-sectional TEM-EDX. Processes occurring on the plasma-facing walls in the Large Helical Device (LHD) and in-vessel material migration were investigated using a technique of material deposition probes. Ten Si plates were located on the outer side of the first-wall surface in each 36° toroidal section (Nos. 1–10). Mixed-material layers deposited on the probes during the fiscal year 2012 plasma campaign contained carbon, boron, hydrogen, and metals from which the plasma-facing elements were made. Metallic impurities and hydrogen content in deposited layers were examined by Rutherford backscattering spectrometry (RBS) and elastic recoil detection analysis (ERDA), respectively. The cross-sectional observations of the deposited mixed-material layers were performed with the help of a scanning electron microscope equipped with an energy dispersive X-ray (EDX) spectrometer. The results of the combined RBS/ERDA analysis and TEM-EDX examination allow tracing the processes occurring on plasma-facing components during the plasma campaign. [ABSTRACT FROM AUTHOR]

Abstract: Samples of polycrystalline ITER-grade W and recrystallized W were irradiated at room temperature with 20MeV W ions to displacement damage up to 0.5dpa. The damaged W samples were then exposed to (i) D neutrals at 403–573K and (ii) D2 gas at 673–1073K and pressures of 1.2 and 100kPa. Trapping of deuterium in the damage zone was examined by the D(3He, p)4He nuclear reaction with 3He energies between 0.69 and 4.0MeV allowing determination of the D concentration up to a depth of 6μm. It has been found that generation of the W-ion-induced displacement damage leads to accumulation of deuterium in the damage zone up to concentration depending on the exposure temperature and, at temperatures ⩾673K, on the D2 gas pressure. Thermal desorption spectra allowed a conclusion that deuterium is mainly accumulated in the form of D atoms bound to inner walls of vacancy clusters. [Copyright &y& Elsevier]

Abstract: As plasma-facing material in the Iter divertor area, tungsten will be subjected to bombardment with low-energy, high-flux deuterium and tritium particles, as well as 14MeV neutrons. At elevated temperatures, hydrogen isotopes can diffuse into W material and can be trapped in n-induced defects. To account for this effect, the trapping of diffusing deuterium at defects generated in the rear side of the 25 and 500μm thick W samples damaged up to a depth of 2.2μm to a damage level of ∼0.3dpa was studied. The front side of each sample was exposed to a deuterium plasma at elevated temperatures. The results show that D atoms permeate through the whole width of the sample and are trapped in ion-induced defects. The highest D retention was measured at 460K – beyond that point, due to the de-trapping process, it decreases with increasing irradiation temperature. [Copyright &y& Elsevier]

Abstract: Deuterium retention in porous vacuum plasma-sprayed tungsten coating has been examined after exposure to low-energy, high-flux (1022 D/m2 s) pure D and helium-seeded D plasmas to ion fluences in the range from 1026 to 1027 D/m2 at various temperatures. The methods used were thermal desorption spectroscopy and the D(3He,p)4He nuclear reaction, allowing determination of the D concentration at depths up to 7μm. Under exposure to pure D plasma (76eV ) at 340–560K, the D concentration reaches 0.1–0.2at.% at depths of several micrometers, while at temperatures above 700K the D concentration is below 10−2 at.% and deuterium is retained over the whole thickness of the coating. Seeding of 76eV He+ into the D plasma reduces the D retention at temperatures of 400–600K. However, at temperatures above 700K, the D retention becomes comparable to that for pure D plasma exposure. [Copyright &y& Elsevier]

Fe–W binary mixed layers were prepared as a model system for reduced-activation ferritic–martensitic (RAFM) steel for studying their dynamic erosion behavior resulting from energetic deuterium (D) irradiation. This investigation aims toward an assessment of RAFM steels as plasma-facing material. The surface composition of the model layers is modified by D irradiation. W is enriched at the surface with D irradiation fluence due to the preferential sputtering of Fe. It depends on the D impinging energy as well as the initial W fraction of the Fe–W layer. No significant development of surface topography was observed within the examined conditions. The erosion yield of a Fe–W layer is comparable to that of pure Fe in the low-fluence range and decreases with increasing D fluence. These results indicate that the dynamic change of erosion yield is significantly correlated with the surface W enrichment. [ABSTRACT FROM AUTHOR]

Abstract: Hydrogen isotope exchange in re-crystallized polycrystalline tungsten was investigated at 320 and 450K. In a first step the tungsten samples were loaded with deuterium to a fluence of 1024 D/m2 from a low-temperature plasma at 200eV/D particle energy. In a second step, H was implanted at the same particle energy and similar target temperature with a mass-separated ion beam at different ion fluences ranging from 2×1020 to 7.5×1023 H/m2. The analytic methods used were nuclear reaction analysis with D(3He,p)α reaction and elastic recoil detection analysis with 4He. In order to determine the D concentration at depths of up to 7.4μm the 3He energy was varied from 0.5 to 4.5MeV. It was found that already at an H fluence of 2×1020 H/m2, i.e. at 1/5000 of the initial D fluence, about 30% of the retained D was released. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface at 320K, but extending to all analyzable depths at 450K especially at high fluences, leading to higher release efficiency. The reverse sequence of hydrogen isotopes allowed the analysis of the replacing isotope and showed that the release of D is balanced by the uptake of H. It also shows that hydrogen does not diffuse through a region of filled traps into a region were unfilled traps can be encounter but transport is rather a dynamic process of trapping and de-trapping even at 320K. Initial D retention in H loaded W is an order of magnitude higher than in pristine W, indicating that every H-containing trap is a potential trap for D. In consequence, hydrogen isotope exchange is not a viable method to significantly enhance the operation time before the tritium inventory limit is reached but should be considered an option to reduce the tritium inventory in ITER before major interventions at the end of an operation period. [Copyright &y& Elsevier]

Abstract: Blistering, deuterium retention and local melting at tungsten exposed to high-fluence (up to 1027 D/m2) of high-flux (1022 D+/m2/s) and low-energy (38eV) deuterium plasma at varying temperature were examined with scanning electron microscopy (SEM), focused ion beam (FIB), thermal desorption spectroscopy (TDS) and electron probe microanalysis (EPMA). Blisters with various shapes were observed after the plasma exposure, and both deuterium retention and blistering showed a significant dependence upon the exposure temperature and fluence. Local melting and flakes on some grains were observed even if the heat flux is less than 0.1MW/m2, suggesting that the local melting is caused by flakes due to the loss of heat conduction between the flakes and the bulk during the plasma exposure. [Copyright &y& Elsevier]

Reduced-activation ferritic–martensitic steels (RAFMS) are advanced structural materials for the construction of future fusion reactors with high fluxes of neutrons, such as DEMO or a Fusion Neutron Source (FNS). In the present work the influence of different damages on deuterium retention in the RAFM Rusfer (Chernov et al., 2007) was investigated. Three different types of damage were applied: (i) irradiation by 20 MeV W 6+ ions to a damage fluence of 0.89 dpa (1.4 × 10 18 ions/m 2 ). Tungsten ion irradiation was used as proxy for displacement damage created by neutrons; (ii) heat loads in the QSPA-T facility with 10 pulses of 0.5 MJ/m 2 with a duration of 0.5 ms; (iii) low-temperature hydrogen plasma irradiation in the LENTA facility at 320 and 600 K to a fluence of 10 25 H/m 2 . The hydrogen isotope retention properties of the damaged and undamaged material were investigated by exposure to deuterium gas several weeks after damaging. The deuterium retention was investigated in the temperature range of RT–773 K at pressure 10 4 Pa. Deuterium depth profiles were measured a month after gas exposure by nuclear reaction analysis (NRA) using the D( 3 He,p)α nuclear reaction. Deuterium retention in damaged and undamaged Rusfer in the temperature range of 300–600 K has a maximum at 500 K for all types of damage investigated. The typical value of deuterium concentration in the bulk is 10 −3 at.%. Peculiarities of D retention in damaged samples are discussed. [ABSTRACT FROM AUTHOR]

Abstract: Deuterium retention in single-crystal tungsten was measured by thermal desorption spectroscopy as a function of D+ energy, fluence and specimen temperature. At 300K, D retention increased significantly with increasing ion energy but at 500K the energy dependence was weak. For both 300 and 500K implantations, the retention was significantly lower for 10eV/D+ than for 500eV/D+ over the fluence range ∼1022–1024 D/m2. Depth profiles in D-implanted SCW show D retention at 300K extending beyond 7μm while at 500K retention concentration is detectable at the surface <60nm. By comparison, D retention of 10eV/D+ at both 300 and 500K is detectable in the very near surface. Similar TDS release peaks for both 10 and 500eV/D+ indicate that trapping mechanisms are independent of incident ion energy, and the reduced D retention is likely to be due to the shallower trapping depth. [Copyright &y& Elsevier]

Abstract: Retention of D in neutron-irradiated W and desorption were examined after plasma exposure at 773K. Deuterium was accumulated at a relatively high concentration up to a large depth of 50–100μm due to the trapping effects of defects uniformly induced in the bulk. A significant D release in a vacuum continued to temperatures⩾1173K because of the small effective diffusion coefficient and the long diffusion distance. Exposure of ion-irradiated W to D2 gas showed a clear correlation between concentrations of trapped and solute D as determined by the trapping–detrapping equilibrium. These observations indicated that the accumulation of tritium in high concentrations is possible even at high temperatures if the concentration of solute tritium is high, and baking at moderate temperatures is ineffective for removal of tritium deeply penetrating into the bulk. Nevertheless, clear enhancement of D release was observed under the presence of solute H. [Copyright &y& Elsevier]

Abstract: The effect of heavy sputtering and of neutron irradiation simulated by displacement damaging with of 20MeV W6+ ions on the optical properties of tungsten mirrors was studied. Ar+ ions with 600eV of energy were used as imitation of charge exchange atoms ejected from fusion plasma. The ion fluence dependence of the surface topography and the optical properties of polycrystalline, recrystallized tungsten (grain size 20–100μm) were studied by optical microscopy, interferometry, reflectometry and ellipsometry. Furthermore, after sputtering in total a layer of 3.9μm in thickness, the orientation and the thickness of the eroded layer of many individual grains was determined by electron backscattering diffraction and confocal laser scanning microscopy. Concluding from the obtained data the neutron irradiation, at least at the damage level would be achieved in ITER, has not to make an additional contribution in the processes developing under impact of charge exchange atoms only. [Copyright &y& Elsevier]

Abstract: The prediction of hydrogen isotope retention in n-irradiated tungsten (W) is an important topic for edge plasma research as it has significant implications for the operation of ITER and fusion reactors. Direct measurement of tritium retention in n-irradiated tungsten is a difficult task. The self-implantation of W is the most promising approach because it generates dense cascade with large clusters which are typical for n-irradiation and any chemical effects are avoided. Pre-damaged samples were exposed to pure and He-seeded deuterium plasmas on both front (undamaged) and rear (damaged) sides in order to separate deuterium (D) retention in plasma-induced traps and in displacement damage created by fast tungsten ions. Irradiation was done up to a damage level expected from the end-of-life neutron fluence in ITER. The diffusion model with dynamic trap formation during irradiation was validated by comparison with experiments and applied for calculation of hydrogen isotope retention in ITER in the presence of neutron irradiation. [Copyright &y& Elsevier]

Abstract: 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 8μm due to the presence of deep (>3.5μm) 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 3.5μm-thick deuterated amorphous carbon layer deposited on the CFC surface. We show that specific localized retention sites correspond to the filling of relatively large (∼3μm.) and deep (at least 3μm below the initial CFC surface) cracks between fibres and matrix by the deuterated amorphous carbon layer. [Copyright &y& Elsevier]

Abstract: Deuterium depth profiles in graphite tiles of JT-60U have been determined by means of the D(3He,p)4He nuclear reaction. Tiles with net carbon deposition showed high deuterium retention. A D/C ratio of 0.07 was found in the outer dome wing tile. In the inner divertor tile covered with a thick deposition layer, the depth profile shows a broad peak in a deep region. This indicates that deuterium trapped in the shallow region can be replaced by hydrogen with hydrogen discharges due to isotope exchange, while deuterium trapped in the deeper region can not. Deuterium in the redeposited layers can be more easily removed than that in the substrate. The deuterium concentration in the outer divertor target tile was very low (D/C<0.003). The highest D/C ratio, 0.09, was found in the upper region of the first wall. [Copyright &y& Elsevier]