Reduced-activation ferritic-martensitic steels (RAFMS) are considered not only as structural but also, under certain conditions, as plasma-facing materials for fusion installations. The base material of RAFMS is iron. These steels also contain 8–12 wt

Тhe results of an experimental study of tungsten as a coating material for the divertor and the first wall of a thermonuclear tokamak reactor are presented. A method for modeling radiation damage of a material by thermonuclear neutrons based on the use of accelerated protons has been developed. A technique for irradiating tungsten samples (PLANSEE) with protons is developed at the cyclotron of the National Research Center Kurchatov Institute. During irradiation, protons with an energy of 3.7 MeV and the total fluence of 10 18 protons/cm 2 are used. The samples irradiated (0.05 dpa) are studied in deuterium plasma on the LENTA device. The results of irradiation and plasma exposure on the surface and the accumulation of deuterium in the irradiated material are presented.

The selective sputtering of reduced-activation ferritic-martensitic steel Eurofer upon deuterium plasma irradiation with ion energy of 100 eV is investigated. The experiments are carried out at 550 K and the radiation-dose range of (3–70) × 1024 ion/m2, and in the temperature range 350–730 K at fixed a radiation dose of 3 × 1025 ion/m2. The sputtering-coefficient dependences and information about the complex relief formed on the surface of Eurofer steel under ion irradiation are obtained. The elemental composition of the surface layer is studied by energy-dispersive X-ray spectroscopy and Rutherford backscattering. Enrichment of the surface layer with tungsten is found, which increases with the fluence of plasma irradiation. At the maximum radiation dose, the concentration of tungsten on the surface, averaged over an area of the analyzing beam of 1.7 mm2, was increased by about 13 times and reached a value of about 6 at % (about 14 wt %).

Results of the complex experimental research of plasma impact on fusion reactor materials are presented. The near-wall plasma of a tokamak reactor is simulated on the linear plasma device LENTA (National Research Center Kurchatov Institute). Plasma fluence of 10 22 –10 23 cm –2 to the material surface is provided at 10 12 –10 13 cm –3 of plasma density in steady-state operation of the device, thus simulating the continuous regime of the fusion reactor plasma-wall conditions. The neutron effect on the first wall material (radiation damage) is also simulated by irradiation with high-energy ions accelerated by a cyclotron to MeV-range energies. The work is centered mainly on tungsten being a candidate for coating of the divertor region in the tokamak reactor. Samples irradiated at doses of 10 21 –10 23 ion/cm 2 to a high damage level from 0.1 to 80–100 displacements per atom characteristic of a durable operation of the reactor have been obtained on the cyclotron at the National Research Center Kurchatov Institute. Helium, carbon, and nitrogen ions and protons whose defect generation mechanisms are very different have been used in irradiations. Erosion data (erosion rate, erosion yield), swelling characteristics (profilometry), and microstructure changes (SEM) of the damaged surface layer are given for tungsten preirradiated with fast nitrogen ions. Proton-irradiated silicon carbide SiC has also been studied in deuterium plasma, and changes in its microstructure are found.

In this paper the results of the studies of reduced activation ferritic-martensitic steel EK-181 (Rusfer-EK-181, Fe–12 Cr–2 W–V–Ta–B) developed in Russia (Bochvar High-Technology Scientific Research Institute of Inorganic Materials—VNIINM) as a structural material for application as the core components of fast neutron, fusion, and hybrid reactors are summarized. Various aspects of hydrogen interaction with Rusfer steel, primarily retention, diffusion, and the effect of various defects on retention, are considered.

At the present time, tungsten is considered as one of the plasma-facing materials of future fusion reactors. As plasma-facing material, W will be subjected to intensive fluxes of deuterium, tritium, helium particles as well as 14 MeV neutrons (n) from the D–T fusion reaction. Neutron irradiation will cause a change in the microstructure of W by creating displacements in the bulk and Re and Os formation [1]. It was shown that the addition of 5 at.% Re to W strongly reduced hydrogen isotopes retention after heavy ion irradiation [2]. It is interesting to study effects of other dopant elements, such as Mo and Ta, on hydrogen isotopes retention. In this study plates of monocrystalline W, W-xTa alloys (x = 1; 3; 5 concentration in at.%) were used as experimental samples. All the samples were polished to a mirror surface and outgassed in vacuum at 1100 K during 2 hours. Sets of W, W-1Ta, W-3Ta, W-5Ta samples were irradiated with low-temperature D plasma up to fluences of 1e24 D/m. Another sets of W, W-1Ta, W-3Ta, W-5Ta samples were exposed in D gas in a temperature range of 400600 K, pressure 10 Pa. The D retention in W and W-Ta alloys was measured by thermal desorption spectroscopy (TDS). An influence of Ta dopant on deuterium retention in W monocrystals after gas exposure and plasma irradiation investigated by TDS is discussed. The results of SEM investigation before and after plasma irradiation are presented.

The effect of deuterium plasma on tungsten with high levels of radiation damage was studied experimentally. Tungsten was examined as a candidate plasma-facing material for a fusion reactor. The effect of damage accumulation in a material irradiated with fusion neutrons was simulated using high-energy ions. Primary radiation defects of 1–100 dpa were produced in tungsten irradiated with He2+ and C3+ ions accelerated to 3–10 MeV in the cyclotron at the Kurchatov Institute with a fluence of 1017–1019 ion/cm2. The irradiated material was exposed to deuterium plasma at the LENTA linear plasma facility operated in the continuous regime and providing a plasma flux of 1021–1022 D/cm2. The erosion dynamics, the development of surface microstructure, and the accumulation of deuterium in tungsten were studied. Enhanced retention of deuterium was observed in the samples damaged both by helium and carbon ions at room temperature (ERDA). The effect of deuterium retention was suppressed in the damaged tungsten samples processed at a temperature of 500°C.

Thermal desorption of deuterium from W was investigated. Virgin samples and samples damaged by 10 MeV C3+ ions were implanted from plasma in the LENTA facility at 370 K and 773 K. In comparison with the undamaged sample, deuterium retention in the damaged sample slightly increased in the case of deuterium implantation at RT, but decreased in the case of deuterium implantation at 773 K. At 773 K, deuterium was concluded to diffuse far behind the D ion range in the virgin sample, while C implantation region was concluded to be a barrier for D diffusion in the damaged sample.

The paper presents an overview of plasma-material interaction studies at the Siberian Synchrotron and Terahertz Radiation Centre. The measurements of recrystallization and surface texturing are demonstrated. The growth of grain size after the exposure by electron beam was detected using two-dimensional diffractometry. The orientation of crystal structure was detected in the tungsten irradiated by deuterium plasma. The residual stresses in irradiated tungsten were measured. The predicted structure of deformations after pulsed heat load is confirmed. First in-situ experiments confirmed the presence of crystal plane rotation effect during a pulsed heat load.

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(He-3, p)He-4 nuclear reaction with 3He energies between 0.69 and 4.5 MeV allowing determination of the D concentration up to a depth of 8 mu m. 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. (C) 2014 Elsevier B.V. All rights reserved.

The paper presents the development of experimental investigations and recent results of the impact on tungsten at high level of radiation damage under steady-state deuterium plasma. Tungsten is considered as a plasma facing material for a fusion reactor. The effect of fusion neutron impact is simulated by surrogate irradiations with high-energy ions. The primary defects at 1-100 dpa were produced in tungsten samples by He and C ions accelerated in the Kurchatov cyclotron to 3-10 MeV at the total fluence of 1017-1019 cm-2. The irradiated material was studied in deuterium plasma on the LENTA linear divertor simulator at the plasma fluence 1021-1022 D/cm2. Erosion dynamics, development of the surface microstructure and deuterium retention were analyzed. Increased deuterium retention detected previously in tungsten pre-irradiated by He ions was also registered (ERDA) on C-irradiated samples at 2-3 dpa. In contrast, a significant decrease in the D uptake has been observed on those samples operated in the experiments at 500°C.

The effects of displacement damage, plasma exposure and heat loads on T retention in reduced-activation ferritic/martensitic (RAFM) steels were investigated by exposing the steels to DT gas at 473 K Despite enormous change in surface morphology, T retention in the heat-loaded specimen was comparable with that in the unloaded specimen. The exposure to plasma resulted in a drastic increase in T retention at the surface and/or sub surface. However, the T trapped at the surface/subsurface was easily removed by maintaining the specimens in air at similar to 300 K Formation of radiation-induced defects led to a significant increase in T retention, and T trapped in the defects was not removed at similar to 300 K These observations suggest that displacement damages have the largest effects on T retention at similar to 473 K

Effect of radiation damage on tungsten response to plasma exposure is examined. Samples of tungsten W 99.95wt% were irradiated with fast 12C3+ ions at 10MeV to fluence of (1–2)×1017ion/cm2 to produce damage in the range 1–40dpa relevant to the level of interest for fusion research. Exposure of the irradiated tungsten to deuterium plasma on linear simulator in erosion condition provided D-ion total fluence (1–3)×1021D+/cm2 at ion energy 250eV. Swelling effect has been observed, and important changes in the structure of the irradiated material have been found (SEM). Erosion of the damaged layer was evaluated at Yd-w≅(3–5)⋅10−3at/ion. The retained deuterium profiles have been taken by ERDA, maximal deuterium concentration reached 6–8at.% in the layer ∼30nm deep corresponding to 2–3dpa. Comparison of D-retention was made for C- and He-irradiated tungsten for this level of damage.

The effects occurring on the surface of tungsten under irradiation with fast ions with an energy in the megaelectrolvolt range and with high fluxes of hydrogen (deuterium) plasma are considered. These effects are radiation damage of the surface layer of the material, its erosion and deuterium retention in it. Irradiation with helium 4He2+ (3.2–4.0 MeV) and carbon 12C3+ (10 MeV) ions is performed using a cyclotron at the National Research Center Kurchatov Institute. The thickness of the damaged layer is 3.5–6 μm. The irradiated samples are exposed to steady-state deuterium plasma using a LENTA linear plasma facility to reach a plasma ion fluence of 1021–1022 cm−2. Tungsten erosion and modification of the structure of the damaged layer are analyzed at a plasma-ion energy of 250 eV. Deuterium retention in the damaged layer is studied by elastic recoil detection analysis. The deuterium concentration and its penetration depth into the material are measured. The data obtained for different kinds of fast ions used in the work are compared.

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 W6+ ions to a damage fluence of 0.89 dpa (1.4 x 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 104 Pa. Deuterium depth profiles were measured a month after gas exposure by nuclear reaction analysis (NRA) using the D(He-3,p)alpha 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. (C) 2014 Elsevier B.V. All rights reserved.

Experimental studies of tungsten (as a candidate plasma-facing material for a fusion reactor) whose properties will degrade as a result of its contact with near-wall plasma and irradiation with neutrons are performed. The effect of a high level of radiation damages (1–100 displacements per atom) on deuterium accumulation and erosion caused by tungsten irradiation with deuterium plasma was studied. Radiation damages are obtained as a result of the irradiation of tungsten samples with high-energy ions in an accelerator (He +2 , C +3 , 4–10 MeV). Then the samples are exposed to steady-state deuterium plasma at the LENTA facility (National Research Centre Kurchatov Institute). The effects of the erosion of tungsten and accumulation of hydrogen isotopes in it are studied. Modification of the surface microstructure and radiation swelling is observed. The helium and deuterium concentrations were measured using the methods of nuclear elastic backscattering and elastic recoil detection analysis. An increased accumulation of deuterium in the damaged layer to a depth of about 5 μm is revealed.

Experimental study aimed at investigation of neutron induced damage influence on fusion reactor plasma facing materials is reported. Displacement damage was produced in tungsten by high-energy helium and carbon ions at 3–10MeV. The reached level of displacement damage ranged from several dpa to 600dpa. The properties of the irradiated tungsten were studied in steady-state deuterium plasma on the LENTA linear divertor simulator. Plasma exposures were made at 250eV of ion energy to fluence 1021–1022ion/сm2. Erosion dynamics of the damaged layer and deuterium retention were observed. Surface microstructure modifications and important damage of the 5μm layer shown. Deuterium retention in helium-damaged tungsten (ERD) showed its complex behavior (increase or decrease) depending on implanted helium quantity and the structure of the surface layer.