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• The irradiated sample with higher W content shows swelling and melting for all discharges, the CrNbTaVW only shows blisters. • The effect of irradiation was more severe in CrNbTaVW 1.7 when compared to CrNbTaVW. • The deuterium retention was evidenced to be in a deeper depth between 1.3 and 3 μm. • Deuterium retention was higher for CrNbTaVW 1.7 when compared with CrNbTaVW for 3 discharges applied. CrNbTaVW x high-entropy alloys have been developed for plasma facing components to be applied in nuclear fusion reactors. The CrNbTaVW x (x = 1 and 1.7) compositions were prepared by ball milling and consolidated at 1600 °C under 90 MPa. To study the irradiation resistance of these materials, deuterium plasmas were used to irradiate the samples in the PF-1000U facility with 1 and 3 discharges. Structural changes before and after irradiation were analyzed by scanning electron microscopy coupled with energy dispersive X-ray spectroscopy. Nuclear reaction analysis was carried out with 1000 and 2300 keV 3He+ ion beams to evaluate the profile and amount of retained deuterium on the irradiated samples. After irradiation, the sample with higher W content revealed swelling and melting for all discharges, while in the case of CrNbTaVW only blisters were observed. The deuterium retention was higher for CrNbTaVW 1.7 when compared with CrNbTaVW for 3 discharges applied. [ABSTRACT FROM AUTHOR]

• Detailed surface and in-depth chemistry of four ITER-like plasma-facing units located in different regions of the divertor, exposed to the WEST tokamak environment after the C4 campaign, were analyzed using X-ray photoelectron spectroscopy. • Boron carbides, tungsten carbides, and tungsten borides dominated the re-deposit region. • A blended mixture of oxidized and metallic tungsten, followed by boron carbides within a 50 nm depth range, was formed in the erosion-dominant area. • The re-deposited mixture in the deposition-dominated area of the divertor, thicker than a hundred nanometers, may significantly modify the hydrogen fuel retention of the plasma-facing units. Four monoblocks coming from one ITER-like plasma-facing unit from the Q3B sector of the lower divertor, named as monoblock (MB)3, MB9, MB20, and MB30, were exposed to the deuterium and helium plasma mixture during the C3 (2018) and C4 (2019) campaigns of the Tungsten Environment in Steady-state Tokamak (WEST), followed by a detailed ex-situ X-ray photoelectron spectroscopy investigation. The surface and in-depth chemistry of the tungsten monoblocks indicated the formation of a re-deposited mixture in the deposition-dominated area of the divertor, thicker than 218 and 172 nm for MB3 and MB9, respectively. The redeposition layer was dominated by a mixture of boron carbides accompanied by tungsten carbides in MB3, while in MB9, the redeposition layer was dominated by tungsten borides. The remaining two monoblocks, MB20 and MB30, were collected from the erosion region and showed similar chemical behavior with a blended mixture of oxidized and metallic tungsten followed by boron carbides within a 50 nm depth range. Boron fixation in the layers is an expected consequence of the boronizations used during the operation, but the chemical status of redeposited elements was characterized for the first time in this work. [ABSTRACT FROM AUTHOR]

Edge engineering is important for both fundamental research and applications as the device size decreases to nanometer scale. This is especially the case for graphene because a graphene edge shows totally different electronic properties depending on the atomic structure and the termination. It has recently been shown that an atomically precise zigzag edge can be obtained by etching graphene and graphite using hydrogen (H) plasma. However, edge termination had not been studied directly. In this study, termination of edges created by H-plasma is studied by high-resolution electron energy loss spectroscopy to show that the edge is sp 2 bonded and the edge carbon atom is terminated by only one H atom. This suggests that an ideal zigzag edge, which is not only atomically precise but also sp 2 bonding, can be obtained by H-plasma etching. Etching of the graphite surface with plasma of a different isotope, deuterium (D), is also studied by scanning tunneling microscopy to show that D-plasma anisotropically etches graphite less efficiently, although it can make defects more efficiently, than H-plasma. [Display omitted] [ABSTRACT FROM AUTHOR]

Surface modification, sputtering erosion and deuterium (D) retention were investigated of tungsten (W) and W-1 wt.% La 2 O 3 (WL10) that were exposed to pure D plasma and D plasma with different impurities (He, N 2 , Ar). After the plasma exposure, the surface of WL10 is covered by blisters with various shapes, and their size and density increase with D fluence. When the fluence was higher than 8.6 × 1024 D m−2, the blister rupturing occurred causing a partially- and/or fully-opened lid features. The phase interface between La 2 O 3 grains and W matrix was served as the nucleation sites for blistering and large-scale crack propagation near the surface. Small blisters were preferentially appeared on the grains with surface orientation of nearly (111) and that originated from the intra-granular crack at depths much closer to the sample surface. Micro-sized pits were formed on the WL10 surface after exposure to high fluence plasma, which was due to a higher sputtering rate of La 2 O 3 grains during plasma exposure. For both materials, seeding He in D plasma can effectively inhibit the blister formation, and most serious blistering occurs with N seeding in D plasma. When exposed to Ar + D plasma, sparse small blisters were visible and most of blisters were ruptured for W, whereas there are some big blisters with long strip shape for WL10. The difference in blistering behavior between W and WL10 was attributed to the differences in microstructure, composition as well as ductility due to the presence of La 2 O 3 phases. The sputtering yields increased slightly with He seeding in D plasma. N 2 and Ar seeding in D plasma leaded to a significant increase in sputtering yields. As compared to the W, the WL10 exhibits a better sputtering erosion resistance in terms of the sputtering yield. The D retention in WL10 increases with D fluence and that was equal to or lower than that in W when exposed to the same conditions. • Both deuterium (D) retention and blistering showed a strong dependence on the exposure fluence and impurity gas. • The phase interfaces between La 2 O 3 and tungsten (W) grains served as the nucleation sites for blistering. • Compared to the W grains, the La 2 O 3 grains were sputtered preferentially during plasma exposure. • The W-1 wt.% La 2 O 3 alloy (WL10) displayed much higher sputtering erosion resistance than pure W. • The amount of D retained in WL10 was dominated by intrinsic bulk defects and that was lower than that in pure W. [ABSTRACT FROM AUTHOR]

Pure tungsten (W) and two representative W alloys of W-1 wt.%Y 2 O 3 and W-0.5 wt.%TiC were subjected to deuterium (D) plasma exposure at a fluence up to 1 × 1026 D/m2 at 400 K to evaluate their surface blistering and D desorption behaviors. Thermal desorption spectroscopy (TDS) results revealed that high-temperature desorption peaks (> 650 K) dominated in HERF-W, whereas the two W alloys featured higher intensities of low-temperature D desorption peaks (< 650 K). Even at a low D implantation fluence of 5 × 1023 D/m2, extensive blisters were evident on the surface of pure W, while no discernible blisters were detected on the surface of W-1 wt.%Y 2 O 3 and W-0.5 wt.%TiC, even with a fluence exceeding 2 × 1025 D/m2. Microstructure analysis indicated that the relatively large grain size and limited intrinsic defects in HERF-W were the primary reasons for the early formation of D-induced blisters and subsequent high-temperature desorption of these trapped D, compared to those of two other W alloys. A focus ion beam (FIB)-SEM technique was employed to explore the mechanisms of blistering. Results revealed that the cavities beneath blisters were primarily located along grain boundaries, and coarse blisters were formed by the merging of several small ones. This work provides insight into the relationships among material microstructures, D-induced blistering and D desorption in W-based materials, thus offering valuable guidance for the design of W materials with high resistance to D-induced damages. [ABSTRACT FROM AUTHOR]

Surface modification, deuterium trapping, microstructure, hardness of tungsten – titanium alloys (W-5Ti) and reference tungsten (W) were investigated. All the specimens were exposed to deuterium plasma (100 eV) for 2 h and 4 h with a density of 1.8 × 1021 D (m−2 s−1) at 470 K. Surface blistering and D retention in tungsten is strongly suppressed by Ti doping. Two typical regions are observed on W-5Ti alloys, namely, W–Ti solid solute area and Ti-enriched region. Most of blisters appeared on Ti-enriched area while no blisters were observed on the W–Ti solid solute region. In addition, the hardness of W is significantly improved by Ti-doping. [ABSTRACT FROM AUTHOR]

A refractory WTaCrVTi high-entropy alloy (HEA) film was prepared by magnetron sputtering, and a pure tungsten (W) film was also prepared in the same way used as a comparison sample. The electrical resistivity of the films in the temperature range of 4 K–300 K was measured by the standard four-probe method and the results showed that the resistivity of the HEA film is about 10 times higher than that of the W film. Furthermore, the thermal conductivity of the film sample is calculated by Wiedemann-Franz law. The results show that the thermal conductivity of the HEA film at 300 K is about 10 times lower than that of the W film. Deuterium and helium retention and desorption behavior induced by deuterium and helium plasma exposure of the two films were investigated. The total deuterium retention measured by thermal desorption spectra (TDS) in the HEA film was approximately 16 times larger than that in the W film, which was attributed to the existence of more grain boundary defects and vacancy-type defects in HEA film, leading to more deuterium trapped. The characteristics of helium bubbles in the two films were compared. The results showed that the helium bubbles in the HEA film had a shallower depth distribution and smaller size than those in the W film, which indicated that the bubbles growth evolution was suppressed in the HEA film compared to the W film. It is suggested that helium atoms in the HEA film experience larger lattice potential energy fluctuations due to the abundant hopping processes during diffusion and aggregation. [ABSTRACT FROM AUTHOR]

Novel highly-oriented (111)-textured boron-doped diamond electrodes (BDD D) featuring high electrochemical activity and electrode stability toward electrochemical analytics were fabricated by deuterium-rich microwave plasma CVD. The high flux deuterium plasma-induced preferential formation of (111)-faceted diamond as revealed by XRD. The highly-oriented diamond surface exhibited improved boron dopant incorporation and activation, whereas the crystals showed enhanced carrier electron acceptance and donation, which accelerated the electron transfer during electrochemical redox mediation. The standard rate constant and peak-to-peak separation ΔE for the oxidation and reduction of the Fe(CN) 6 3−/4- and Ru(NH 3) 6 2+/3+ redox probes reached ΔE values of only 60.6 and 59.8 mV, respectively. The enriched electrochemical performance of the BDD D electrodes is an advantageous feature allowing them to be applied as ultrasensitive electrodes, demonstrated here by paracetamol determination. The differential pulse voltammetry results revealed an enhanced electrochemical oxidation effect for paracetamol at the deuterium-grown (111)-rich diamond electrode. A single linear range from 1 to 125 μM along with a low detection limit of 0.76 μM were achieved. Image 1 • Novel deuterium-plasma-grown diamond electrodes were prepared. • Deuterium enriched electrode exhibited high electroactivity and stability. • Enhanced electrochemical mechanism of the diamond electrodes was elaborated. • Wide linear range of sensing of paracetamol from 1 to 125 μM was achieved. [ABSTRACT FROM AUTHOR]

W-SiC composite material is a promising plasma-facing material candidate alternative to pure W due to the low neutron activation, low impurity radiation, and low tritium diffusivity of SiC while leveraging the high erosion resistance of the W armor. Additionally, W and SiC have high thermomechanical compatibility given their similar thermal expansion rates. The present study addresses the synthesis and performance of compositionally graded W-SiC films fabricated by pulsed-DC magnetron sputtering. Compositional gradients were characterized using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy (EDS), and crystallographic information was obtained using electron diffraction and X-ray diffraction (XRD). Samples were exposed to L-mode deuterium plasma discharges in the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Post-mortem characterizations were performed using scanning electron microscopy (SEM) and XRD. Electron diffraction and XRD showed that the compositionally graded W-SiC films were composed of polycrystalline W and amorphous SiC with amorphous W+SiC interlayers. No macroscopic delamination or microstructural changes were observed under mild exposure conditions. This study serves as a preliminary examination of W-SiC compositionally graded composites as a potential candidate divertor material in future tokamak devices. [ABSTRACT FROM AUTHOR]

• Some silicide materials are found out as insert materials for the fabrication of W/CFC blocks. WSi 2 do not exceed but are the same level as insert material as that of SiC + additives. • Zr and V behave as insert materials for W/CFC bonding, but the thermal stability of specimens is not enough. • The bonding strength near edge of an expected size of mono-block is not enough. It needs to be fabricated as a larger size block with 40 MPa of pressure and removed its near-edge parts. The construction of a satellite tokamak, JT-60SA, was completed in 2020. JT-60SA is expected to contribute the research of ITER and developments for DEMO. JT-60SA employs carbon and carbon fiber strengthened carbon (CFC) composites divertors. The divertors will be set on an actively cooled cassette and be remotely handled because of human access restriction into the vacuum vessel due to the radioactivation caused by repeated deuterium plasma operations. An important mission of JT-60SA is high-β steady state plasma research, and during the mission, the use of carbon divertors will be continued. The possibility of metal wall experiments after the complete of high-β steady state plasma research is under discussion. Then the use of remote handled divertor cassette will be a condition, therefore, high performance and appropriate weight divertors are convenient for the operation. One idea is to use clad plates of tungsten and CFC. In previous researches revealed that a sinter bonding method using an insert material of SiC powders + additives was able to join W and CFC. But the research accidentally found out the insert material without explore of the other materials and the size of specimens was small. In present research, some candidates for the insert material were investigated and also the uniformity of strength on a W/CFC block of the expected size were evaluated. [ABSTRACT FROM AUTHOR]

Tungsten (W) film was prepared by magnetron sputtering, and the coarse crystalline W bulk was used as reference samples. All the samples were exposed to deuterium plasma with the exposure fluence of 1.67 × 1024 D/m2. The films were annealed in-situ at 723 K during deposition and hillocks were found on all film surfaces. Focused ion beam analysis of the subsurface structure of hillocks provides data support for the growth mechanism of hillocks. The formation of hillocks suggested that the underlying force is released the intrinsic compressive stress caused by thermal cycling during the magnetron sputtering. After deuterium plasma exposure, the W bulk showed obvious stepped high-dome deuterium blisters. The micrographs revealed the microscopic morphology characteristics of deuterium blisters and hillocks, and analyzed the similarities and differences between hillocks and surface deuterium blisters. By comparing the growth mechanism of hillocks on the W film with the surface blistering on the W bulk, the stress growth model of deuterium blisters on the surface of coarse crystalline W bulk is proposed. This work provides a new perspective for the transverse stress model of deuterium-induced surface blistering. [ABSTRACT FROM AUTHOR]

• We studied the erosion (and deposition) effect on the SiC mirror sample caused by high-energy neutral particles for the expected environmental condition of the field mirror of ITER divertor VUV spectrometer. • By ZEMAX software with a simplified entrance duct model, the flux of neutral particles reaching the first mirror (the field mirror) was calculated based on the particle flux in the first wall. The total depth of erosion on the first mirror of ITER divertor VUV spectrometer was calculated to be around 9 nm using the input particle flux from SOLPS-ITER code. • The erosion resistance test was performed on the SiC mirror sample using the laboratory-scale plasma device to assess the mirror surface damage by erosion from the charge exchange neutral particles of ITER. • In the experiment, the SiC mirror sample was exposed to H + and D + plasmas upto the total erosion depth of 650 nm, and this value is larger than the expected total erosion depth in the first mirror. The surface roughness change of Δ ∼ 0.7 nm was measured, and this mirror degradation is within the allowable level considering the expected initial mirror roughness of the first mirror ∼ 1 nm. A series of tests have been performed to validate the resistance of the Silicon Carbide (SiC) mirror as the first mirror material of ITER VUV spectrometers to all ITER environmental conditions. In the present paper, we focused on the erosion (and deposition) of the SiC mirror sample caused by high-energy neutral particles. The flux of neutral particles reaching the first mirror was calculated using the ZEMAX software with a simplified entrance duct model. In the calculation, the particle flux reaching the first wall is necessary and the previously reported values derived from the SOLPS calculations were used. Based on this estimated particle flux, the erosion resistance tests were performed to check the erosion effect due to the high-energy neutral particles on the first mirror of ITER divertor VUV spectrometer. In the experiment, erosion was induced by exposing the SiC mirror sample to hydrogen and deuterium plasmas (and ions). The target fluence of incident ions in experiment is based on the estimation of the flux of neutral particles in ITER. The surface shape, composition, erosion depth, and surface roughness were measured to check the damage of the mirror surface after erosion test. Based on simulations and erosion resistance tests, it was concluded that the SiC mirror can be used as the first mirror of ITER divertor VUV spectrometer. [ABSTRACT FROM AUTHOR]

• The absence of an oxide layer on the downstream surface of the membrane enhances the plasma-driven permeation flux. • The plasma incident flux is a more crucial factor in enhancing the plasma-driven permeation flux than temperature. • The maximum plasma-driven permeation flux reaches 4.54 × 1018 D/(m2 s) at a radio frequency power of 500 W. The process of particle exhaust based on the cryopump in fusion reactor presents a severe issue of deuterium and tritium retention. However, the metal foil permeation pump represents an effective means to solve the problem of deuterium and tritium retention during the particle exhaust process. How to enhance the permeability performance of metal films to meet the requirements of practical engineering applications of fusion reactor is the main content of the research. This article focuses on the simulation and experimental investigation of V membrane, aiming to explore methods for enhancing membrane permeation flux. Through simulation analysis, the influence of thickness, membrane surface state, and temperature on both plasma-driven and gas-driven permeation has been studied. Among these, the downstream surface condition is changed during the PDP process, resulting in an almost twofold increase in permeation flux. Based on the permeation experiments the influence of temperature on permeation was proved and the impact of plasma source power on permeation was also studied. It was found that the permeation flux increased more than two orders of magnitude when the radio frequency power changed from 0 to 500 W, which indicates that plasma incident flux serves as a more effective means to enhance permeation flux compared to temperature. [ABSTRACT FROM AUTHOR]

Using transient grating spectroscopy, we measure thermal diffusivity and surface acoustic wave speed in tungsten exposed to different fluences of deuterium plasma. Scanning electron microscopy (SEM) shows the formation of surface blisters that have similar morphology for all fluences considered. A significant reduction in thermal diffusivity and surface acoustic wave speed occurs as a result of plasma exposure. A saturation of the thermal diffusivity reduction with fluence is seen. Deuterium ion flux density appears to play a more important role in thermal diffusivity reduction than exposure time. These observations have important implications for plasma facing components in future fusion reactors. Unlabelled Image [ABSTRACT FROM AUTHOR]

• In the Large Helical Device (LHD) of the National Institute for Fusion Science, the experiments using deuterium plasma were performed from March 2017. • The evaluation of the neutron flux in the torus hall as well as the LHD body is very important for designing the decommissioning of the LHD in the future, and the radiation shielding for workers and components around the LHD. • In this study, we performed the activation experiments using multiple activation foils and experimentally obtained the neutron spectrum in the vacuum vessel of the LHD for the first time. In the Large Helical Device (LHD) of the National Institute for Fusion Science, the experiments using deuterium plasma were performed from March 2017. Neutrons with the energy of 2.45 MeV are generated by d (d, n) 3He reactions. The evaluation of the neutron flux in the torus hall as well as the LHD vessel is very important for designing the decommissioning of the LHD in the future, and the radiation shielding for workers and components around LHD. In this study, we performed the activation experiments using multiple activation foils to obtain a neutron spectrum in the vacuum vessel of the LHD. The reaction rate of each activation foil was generally consistent with the calculation result by MCNP6. As a result of the unfolding with SAND-II code, we experimentally obtained the neutron spectrum in the vacuum vessel of the LHD for the first time. [ABSTRACT FROM AUTHOR]

This work aims to investigate the effects of 800 keV argon ions pre-damage on the following deuterium plasma irradiation behavior in tungsten-tantalum alloy (5 wt% tantalum dopant). The un-damaged and pre-damaged W–5%Ta were exposed to 60 eV deuterium plasma at a fluence of 1.944 × 1026 D/m2 in the linear plasma device. The results suggest that pre-damaged induced by Ar8+ ions can effectively mitigate deuterium-induced blistering, while the deuterium retention is significantly higher as compared to the un-damaged W–5%Ta. According to the transmission electron microscopic (TEM) analysis, a large number of defects can be observed clearly after irradiation by 800 keV Ar8+ ions, such as vacancies and dislocation loops, which eventually affect the surface blistering and deuterium retention in W–5%Ta. Furthermore, we also compared the deuterium retention in pre-damaged pure W, which was significantly increased by a factor of 15 than in pre-damaged W–5%Ta. • Pre-damages induced by Ar8+ ions can effectively mitigate deuterium-induced blistering in W-5%Ta. • Higher deuterium retention within the range of 300–1200 K in the pre-damaged W-5%Ta. • The damages induced by Ar8+ ions can lead to the irradiation harden of W-5%Ta. • The deuterium retention in pre-damaged pure W was increased by a factor of 15 than in pre-damaged W-5%Ta. [ABSTRACT FROM AUTHOR]

Tungsten and tungsten-tantalum alloy with tantalum concentration of 10 wt%, sintered by spark plasma sintering, were exposed to deuterium plasma at ∼500 K with flux of 1 × 1021–1 × 1022 D m−2s−1, an ion energy of 40 eV and fluence of 1 × 1025–1 × 1026 D m−2. After exposure, a large number of blisters with an average size of 1–2 μm form on the surface of tungsten, and the area density of blisters increases with increasing irradiation fluence. However, the dominant morphology on the surface of tungsten-tantalum alloy is characterized by the striped microstructure, and exhibits a strong grain-orientation dependence. TDS results show two major desorption peaks located at ∼650 K and ∼770 K for tungsten and tungsten-tantalum alloy, respectively. Furthermore, the deuterium retention in tungsten is 4–6 times more than that in the tungsten-tantalum alloy. It is suggested that tantalum alloying can be expected to suppress the surface blistering and reduce the deuterium retention in tungsten. [ABSTRACT FROM AUTHOR]

Surface morphology and deuterium retention in tungsten have been examined after exposure to a low-energy deuterium plasma with an ion fluence of 6.0 × 1024 D/m2 at various temperatures. Optical differential interference contrast microscope and scanning electron microscope equipped with focused ion beam were applied for surface observations. D (3He, p) 4He nuclear reaction at 3He energies varying from 0.5 to 4.5 MeV was used to determine the deuterium depth distributions within the depth of 8.0 μm. The thermal desorption spectroscopy was applied for deuterium retention measurement. The observed surface blistering and deuterium retention are found to strongly depend on the exposure temperature. At 230 K, many small blisters are formed on the tungsten surface. As the exposure temperature increases to 370 K, the blisters become sparser and larger. At 520 K no blisters appear. The deuterium retention is 1.4 × 1020 D/m2 at 230 K. Then it increases to 1.1 × 1021 D/m2 as the exposure temperature rises to 370 K, and decreases to 2.3 × 1020 D/m2 at 520 K. [ABSTRACT FROM AUTHOR]

In this study the fuel retention and erosion of different reduced-activation ferritic martensitic (RAFM) steels exposed to deuterium (D) plasma were investigated. D plasma exposure was performed in a linear experimental plasma system (LEPS) with the incident ion energy between 30 and 180 eV per D, fluence from 1023 to 1025 D/m2 and sample temperature between 300 and 500 K. After D plasma exposure, erosion structure development of surface topography was observed, but no significant difference was seen among various RAFM steels within the examined conditions. W-enriched surface is formed for all investigated RAFM steels with little difference. RAFM steels exhibited similarities in D erosion and dependence on D plasma energy or fluence. Thermal desorption spectroscopy reveals in all steel samples only one broaden releasing peak from 450 to 800 K and total D inventory of 1018–1020 D/m2. Similar D retention behavior of five RAFM steels was found that it decreases with increasing incident fluence or ion energy. Moreover, different cutting treatment along rolling direction leads to different erosion structure and D releasing behaviors. [ABSTRACT FROM AUTHOR]

The boron-doped diamond thin films were growth in deuterium rich microwave plasma in CVD process. The mechanism of influence of plasma composition on boron doping level was studied using optical emission spectroscopy. Deuterium rich plasma results in an increased dissociation of B 2 H 6 precursor and intense boron-radicals' production. In consequence, a higher doping level of diamond films was observed by means of Laser Induced Breakdown Spectroscopy and Raman spectroscopy. Deuterium species modify the mechanism of boron incorporation into thin films leading to increased boron concentration. Lower concentrations of sp 2 phases and CH defects have been noticed in the films deposited in the plasma with deuterium than with hydrogen. Moreover, all BH x (x = 0–3) species can bind to radical sites on the diamond {100} surface to form stable complexes. The enhanced boron doping is attributed to the lower energy of B H bond at the growth surface, when compared to C H bond. The hydrogen abstraction from B H site is providing dangling bond for diamond growth and boron incorporation. This effect plays a main role due to extended dissociation caused by deuterium rich plasma. The increase in the carrier concentration and the decrease in the Hall mobility for the BDD samples grown in deuterium was registered. Unlabelled Image • Boron-doped diamond films were grown in deuterium rich plasma CVD. • D 2 rich microwave plasma leads to enhanced boron doping. • Lower concentrations of sp 2 phases and CH defects have been noticed in deuterium grown diamond. • Hydrogen abstraction from B H site is enhanced by deuterium rich plasma providing dangling bond for diamond growth. [ABSTRACT FROM AUTHOR]

Abstract Deuterium (D) plasma-driven permeation (PDP) experiments for a Chinese reduced activation martensitic/ferritic steel CLF-1 have been performed. The effects of implantation flux, sample temperature and ion incident energy have been studied. The steady state PDP flux is found to be proportional to the square root of the implantation flux, indicating that the permeation takes place in recombination-limited regime for the upstream surface and diffusion-limited regime for the downstream surface (RD). The PDP flux and flux ratio (PDP flux/implantation flux) decrease with the increase of ion incident energy, which may be due to the enhancement of D recombination coefficient at the upstream surface. [ABSTRACT FROM AUTHOR]

This paper is devoted to experiments on testing of the Sn-Li capillary porous system (CPS) under conditions of deuterium plasma irradiation, carried out at a plasma-beam installation (PBI). As a metal CPS matrix, molybdenum mesh was used. The paper presents a detailed description of the development of technology and procedure for the fabrication of the investigated sample, intended for plasma testing, as well as experiments on irradiation of Sn-Li CPS with deuterium plasma. As a result of the experiments performed, the dependences of the temperature of the investigated sample on the energy of the plasma contacting with the CPS, time dependence of the gas phase composition in the PBI chamber during irradiation of Sn-Li CPS with deuterium plasma, optical spectra depending on the temperature were obtained. Post-experimental material science studies with a sample of Sn-Li CPS were also performed and the results of microstructure, thermal and X-ray phase analysis were obtained. Analysis of the experimental data obtained showed that the energy of deuterium plasma precipitated on the Sn-Li alloy is re-emitted into the energy of optical radiation on the evaporating neutral atoms of lithium and tin. A complete evaporation of the alloy from CPS, during interacting with deuterium plasma, leads to partial destruction of the metal matrix of molybdenum. [ABSTRACT FROM AUTHOR]

Abstract Targets of Reduced Activation Ferritic Martensitic (RAFM) steel F82H were exposed to low-energy (200 eV), high flux (about 1022 D/m2s) deuterium (D) plasma at 623–773 K to various D fluences in the range from 1 × 1025 to 2.5 × 1026 D/m2. The surface morphology of the plasma-exposed targets was examined with a field-emission scanning electron microscope. Cross-sectional observations of nano-structures formed on the F82H target surfaces were performed using a transmission electron microscope equipped with an energy dispersive X-ray spectrometer. It has been shown that nano-sized fiber-like layers are formed on the target surfaces under D plasma exposure. Micro-sized surface morphology pattern depends on the D fluence. As the D fluence increases, clusters of the fiber-like layers begin to be formed and organized into ordered structure. Highlights • F82H steel targets were exposed to 200 eV D plasma at elevated temperatures to various D fluences. • Nano-sized fiber-like layers were formed on the target surfaces. • As the D fluence increased, clusters of the fiber-like layers began to be formed and organized into ordered structure. [ABSTRACT FROM AUTHOR]

Abstract Penning-type ion source performance for neutron generator applications is characterized partly by the atomic ion fraction, providing one path by which source performance can be improved for increased neutron yields. A miniature penning ion source has been fabricated to investigate the atomic ion fraction of deuterium plasma by a mass-energy analyzer. The discharge current and atomic ion fraction increase with the increase of anode voltage and pressure. Effects of electrode materials on discharge characteristics have been studied. The atomic ion fraction could increase about 2% of the original atomic fraction by Au cathode instead of steel cathode, and it could increase about 1% of the original atomic fraction by Al cathode. The results can provide useful information for improving source performance by selecting more suitable anode voltage, pressure and electrode material. [ABSTRACT FROM AUTHOR]

W-Y 2 O 3 alloy was processed by high energy rate forging at 1823 K. Irradiation resistance of the well-developed W-Y 2 O 3 alloy against 50 eV deuterium (D) ions with total fluences ranging from 2.2 × 10 25 D·m −2 to 1.3 × 10 26 D·m −2 was performed in a linear plasma device. During irradiation, the ion flux was up to 1.2 × 10 22 D·m −2 ·s −1 and the temperatures of the samples were kept at 443–463 K. Both the surfaces of perpendicular to (Plane-A) and parallel to (Plane-B) the forging axis were irradiated. After irradiation, D retentions of the irradiated samples were measured by thermal desorption spectroscopy (TDS). The experimental results show that a large number of small-blisters (≤3 μm) were formed on the irradiated Plane-A while hardly formed on the other. However, D retention of the irradiated Plane-B is larger than that of the irradiated Plane-A. The underlying mechanisms are discussed. [ABSTRACT FROM AUTHOR]

Surface blistering and deuterium (D) retention of heavy ion pre-irradiated (1 dpa) tungsten (W) exposed to low-energy (40 eV) and high-flux (1–2 × 10 22 D/m 2 s) D plasma has been investigated with low fluence of 0.1 × 10 27 D/m 2 and a high fluence of 2.2 × 10 27 D/m 2 . Surface morphology observations show that a large number of blisters are formed on the undamaged W after low-fluence exposure while the area density of the blister will have significantly increased and blister bursting will be triggered in the case of high-fluence exposure. In contrast, the heavy ion pre-irradiation noticeably reduces the area density of blisters for both low- and high-fluence exposures. The thermal desorption spectroscopy (TDS) shows that the total D retention in the pre-damaged W is greater than that of the undamaged sample, and this trend is more significant with increasing fluence/duration of D plasma exposure. The results of positron annihilation Doppler broadening spectrometry (PA-DBS) and TDS indicate that a large number of vacancy-type defects, especially those with a higher trapping energy of D, are induced by the heavy ion pre-irradiation. The increasing defects/D-trap sites may result in two outcomes, the D concentration will disperse at each blister nucleation site and D inward diffusion is enhanced and, therefore, lead to the mitigation of D-induced blistering and increase the D retention. In addition, combined with the enhanced D inward diffusion caused by increasing exposure duration, the D retention is therefore much higher in the pre-damaged W in the case of high-fluence exposure. [ABSTRACT FROM AUTHOR]

Reduction of hydrogen content in deuterium-fueled fusion plasmas is important not only to avoid diluting reacting core deuterons but also so as to allow the optimization of hydrogen-minority-heating efficiency in those fusion devices which employ ion-cyclotron-radio-frequency heating systems. In EAST, the amount of hydrogen released from plasma-facing components has been shown to depend strongly on both their composition and their temperature. As measured by thermal desorption spectroscopy, the hydrogen inventory in graphite – used in EAST as lower divertor material – has been determined to be >25 times larger than that of tungsten which comprises the upper divertor. This difference in hydrogen inventory is attributed mostly to the intrinsically porous nature of bulk graphite. Thus the main source of hydrogen release into EAST discharges was identified as the graphite tiles used in the lower divertor. The hydrogen content in EAST plasmas were clearly reduced by first employing a high-temperature vacuum baking of all graphite tiles and then renewing a 100–200 μm thick SiC coating before an EAST experimental run campaign. Subsequent active surface conditioning of all wall components with elemental silicon and then with elemental lithium were seem to again reduce the plasma hydrogen content significantly – with lithium proving to be more effective than silicon. Combining these several techniques, H/(H + D) levels as low as ∼3% have been achieved in EAST discharges. Additionally, the effects of lithium thickness on H surface implantation and retention has been re-examined semi-quantitatively using data from a previous run campaign. These data suggest that relatively thick Li films coated on the first wall can effectively isolate the rich source of hydrogen stored in the porous bulk of the underlying graphite from the deuterium-fueled plasma so as to minimize hydrogen release. Finally an operational maneuver whereby a diverted plasma is repetitively switched from an upper single null configuration to a lower single null configuration is presented. This switching maneuver has been shown to suppress hydrogen influx into a 35s-long EAST discharge by alternately mitigating the rise in divertor temperature. [ABSTRACT FROM AUTHOR]

Tungsten (W) layer deposited in argon and deuterium atmosphere by magnetron sputtering was used as a model system to study the deuterium (D) retention and release behavior in co-deposited W layer. After deposition several selected samples were exposed in deuterium plasma at 370 K with a flux of 4.0 × 10 21 D/(m 2 s) up to a fluence of 1.1 × 10 25 D/m 2 . Structures of co-deposited W layers are investigated by field-emission scanning electron microscopy (FESEM) and X-ray diffraction (XRD), and the corresponding D retention and release behaviors are studied as functions of deposition and exposure parameters using thermal desorption spectroscopy (TDS). Two main D release peaks were detected from TDS spectra located near 600 and 800 K in these W and D co-deposited layers, and total deuterium retention increased linearly as a function of W layer's thickness. After deuterium plasma exposure, the total D retention amount in W layer increases significantly and D release peak shifts to lower temperature. Clearly, despite the high density of defects expected in co-deposited W layers, the initial deuterium retention before exposure to the deuterium plasma is low even for the samples with a W&D layer. But due to the high densities of defects, during the deuterium plasma exposure the deuterium retention increases faster for co-deposited layer than for the bulk W sample. [ABSTRACT FROM AUTHOR]

Surface topography of and deuterium (D) retention in reduced activation ferritic-martensitic Eurofer’97 and ferritic oxide dispersion strengthening ODS-16Cr steels have been studied after exposure at 600 K to low-energy (70 and 200 eV), high-flux (∼10 22 D/m 2 s) pure D and D-10%He plasmas with D fluence of 2 × 10 25 D/m 2 . The methods used were scanning electron microscopy, energy-scanning D( 3 He,p) 4 He nuclear reaction, and thermal desorption spectroscopy. As a result of the plasma exposures, nano-sized structures are formed on the steel surfaces. After exposure to pure D plasmas, a significant fraction of D is accumulated in the bulk, at depths larger than 8 μm. After exposures to D-He plasmas, D is retained mainly in the near-surface layers. In spite of the fact that the He fluence was lower than the D fluence, the He retention in the steels is one order of magnitude higher than the D retention. [ABSTRACT FROM AUTHOR]

Effects of pre-damage by 500 keV argon ion implantation on deuterium-induced blistering in tungsten has been investigated. After low-energy (40 eV) and high-flux (∼10 24 D/m 2 s) deuterium plasma exposure with short exposure duration (100 s), a large increase of deuterium retention is found in the pre-damaged tungsten, while surface blistering is significantly suppressed as compared to the un-damaged one. According to elastic recoil detection analysis, a local deuterium concentration peak is observed at a depth of ∼100 nm for the un-damaged tungsten, which is supposed to be related to the surface blistering with nanometer size. Comparison of deuterium retention in the near surface (within 300 nm) and in the bulk suggests that deuterium inward diffusion is more significant in the pre-damaged tungsten. It is speculated that the creation of deuterium trap-sites and enhancement of deuterium inward diffusion give rise to an increase of critical deuterium concentration for blistering and contribute to the suppressed deuterium-induced blistering on pre-damaged tungsten under the present exposure conditions. [ABSTRACT FROM AUTHOR]

The behavior of tungsten-vanadium (W-V) alloys fabricated by powder metallurgy as a plasma facing material has been studied. W-V alloys with different vanadium concentrations (5 and 10 wt %) manufactured by hot pressing (HP) were exposed to deuterium plasma (flux ∼4.6 × 10 21  m −2 s −1 , fluence ∼5.6 × 10 25  m −2 , ion energy ∼60 eV, target temperature ∼450 K) in the linear plasma device STEP at Beihang University. Three typical grains are observed on HP sintered W-V alloys and exhibit a significant effect on its performance under deuterium plasma irradiation. Surface blistering only occurs at W-enriched grains and is significantly mitigated in W-V alloys, especially in W-10 V, blistering is completely suppressed. On the other hand, deuterium retention dramatically increases in the W-V alloys due to vanadium addition. The deuterium retention in W-5 wt. % V is about 6.2 times more than that in rolled pure W, and this factor further increases to 6.9 when the V concentration rises to 10 wt %. We ascribe these phenomena to the changes of microstructures and components caused by vanadium addition. [ABSTRACT FROM AUTHOR]

In the article, a hydrogen exchange between the Zr–1%Nb alloy (E110) and the gas ambient was experimentally studied when the samples were irradiated with deuterium and argon plasma ions. It has been established that, upon irradiation with deuterium plasma ions with an energy of E = 650 eV/at, the enhanced absorption of deuterium exceeds the release of hydrogen initially contained in the samples, which leads to their loading with hydrogen isotopes. Adding 30 at.% oxygen to the plasma-forming gas or raising the sample temperature from T = 450 K to T = 600 K, significantly reduces the content of hydrogen isotopes in the sample. Based on the aggregate data obtained by atomic and ion irradiation, a mechanism of hydrogen exchange between zirconium alloy and gas ambient is proposed. The process includes three stages: reactions on the oxidized surface of the zirconium alloy (surface hydroxylation and formation of water molecules); reactions at the metal-oxide interface; transfer of hydrogen isotopes through the surface oxide layer in both directions due to hopping between neighboring oxygen ions. Surface reactions caused by irradiation of atoms and ions trigger the hydrogen exchange. The proposed model agrees with the experimental data on the irradiation of the E110 alloy with atoms and ions of hydrogen isotopes. • Plasma ion irradiation causes hydrogen exchange between zirconium alloy and a gaseous ambient. • Irradiation conditions are found to rise or reduce the hydrogen content in zirconium. • A mechanism of the hydrogen exchange under atomic or ion irradiation is formulated. • Surface reactions caused by irradiating species play a key role in the process. [ABSTRACT FROM AUTHOR]

Li-D films co-deposited from deuterium plasma were investigated. It was found that the D concentration has the dependence on the deposition temperature with a maximum of ∼30 at. % at 350–500 K. Lithium mass deposited on the substrate decreases rapidly at deposition temperatures above 570 K. In the narrow range of 520–570 K lithium collection is still efficient and deuterium accumulation is already low, so the temperatures in the range are optimal for lithium collectors in fusion devices. Deuterium in the films is mainly in the form of lithium deuteride with small amounts of D is solution and in lithium hydroxide and lithium hydronitrides. [ABSTRACT FROM AUTHOR]

Hot-rolled tungsten samples were first irradiated with helium ions with varying fluences or energies and then exposed to deuterium plasmas, and the influence of helium fluence (corresponding to damage levels of 0.12–2.3 dpa) and helium ion incident energy (50, 200, 400 keV) on deuterium retention were investigated. A significant increase of the total deuterium inventory was observed in the helium irradiated sample, and deuterium retention reaches saturation at 0.69–1.15 dpa. Radiation-induced defects and helium-vacancy complexes or helium bubbles contribute to the increase of deuterium inventory, and the effect of helium-vacancy complexes or helium bubbles is aggrandizement with the increase of damage level. Comparison of deuterium depth profiles of samples irradiated by helium ions with different incident energies reveals that the vast majority of deuterium is trapped in the helium accumulation region, and the depth of the maximum deuterium concentrations is consistent with the maximum helium concentrations rather than the damage levels. Furthermore, the total amount of deuterium increases with the increase of helium ions energy due to the deepening of the irradiation region. [ABSTRACT FROM AUTHOR]

We characterize the neutron output of a deuterium–deuterium plasma fusion neutron generator, model 35-DD-W-S, manufactured by NSD/Gradel-Fusion. The measured energy spectrum is found to be dominated by neutron peaks at 2.2 MeV and 2.7 MeV. A detailed GEANT4 simulation accurately reproduces the measured energy spectrum and confirms our understanding of the fusion process in this generator. Additionally, a contribution of 14 . 1 MeV neutrons from deuterium–tritium fusion is found at a level of 3 . 5 % , from tritium produced in previous deuterium–deuterium reactions. We have measured both the absolute neutron flux as well as its relative variation on the operational parameters of the generator. We find the flux to be proportional to voltage V 3 . 32 ± 0 . 14 and current I 0 . 97 ± 0 . 01 . Further, we have measured the angular dependence of the neutron emission with respect to the polar angle. We conclude that it is well described by isotropic production of neutrons within the cathode field cage. [ABSTRACT FROM AUTHOR]

One of the most important changes of plasma-facing material is plasma-induced damage which limits the material lifetime in fusion reactors. In this study, the influence of exposure conditions on tungsten nanostructure formation and surface modification was investigated. TEM specimens and bulk samples were prepared and exposed to deuterium plasmas with different fluxes (4.0–10.0 × 10 21 D m −2 s −1 ), fluences (2.5–10.0 × 10 25 D m −2 ) and temperatures (343–513 K), while ion energy was constant (60 eV). Distortion, nano-polycrystalline and amorphous structure were observed as irradiation damage. Increasing exposure intensity results in severer damage: (1) amorphous structure arises with increasing temperature; (2) deformation region greatly extends at a higher fluence; (3) degree of lattice distortion increases in high flux. Furthermore, surface morphology featured by blisters changes dramatically as well. Nanostructures and surface morphology observation indicates that nanostructure evolution is strongly related to surface modification, and it would influence the properties of tungsten. [ABSTRACT FROM AUTHOR]

The TCV tokamak contributes to physics understanding in fusion reactor research by harnessing a wide experimental tool set: in particular flexible shaping and high power electron cyclotron heating. Plasma regimes with high plasma pressure, a wider range of temperature ratios and significant fast-ion population are now attainable with a TCV heating system upgrade. In a first stage, a 1 MW neutral beam was installed (2015) and is reported in this paper. The installation of the NB injector required modifications of the vacuum vessel and considerable work on the machine infrastructure, resulting in a shutdown from late 2013 to mid-2015. TCV is now operating partly as a European Medium-Size Tokamak (MST) facility under the auspices of the EUROfusion consortium. The NBI was intensively operated in the February–July 2016 phase of the MST campaign. Record ion temperatures of 2.0–2.5 keV and toroidal rotation velocities up to 160 km/s were promptly attained in the first few L-mode discharges with NB injection. Ion temperatures up to 3.5 keV were subsequently achieved in ELMy H-mode. The injector produces a focused deuterium neutral beam with 25 keV energy, 1 MW neutral power and 2 s duration. [ABSTRACT FROM AUTHOR]

The effect of the oxidation conditions on hydrogen diffusion in oxide layers in Zr alloy was experimentally studied. Samples of GNF-Ziron (high-iron Zircaloy-2-type alloy) sheets were oxidized in air or steam and then exposed to deuterium plasma to charge them with deuterium. Deuterium depth profiles in each sample were observed by nuclear reaction analysis during plasma exposure, and the deuterium diffusion coefficient D was estimated. The results showed that the D values of the steam-oxidized samples were high, whereas that of a steam-oxidized sample that was subsequently heated in vacuum was as low as those of air-oxidized samples. The difference in D was not attributed to the structure but is probably due to the presence of OH groups absorbed during steam oxidation. [ABSTRACT FROM AUTHOR]

In this work, tungsten nitride coatings with nitrogen content in the range of 19–50 at% were prepared by reactive multi-pulse high power impulse magnetron sputtering as a function of the argon and nitrogen mixture and further exposed to a deuterium plasma jet. The elemental composition, morphological properties and physical structure of the samples were investigated by Rutherford backscattering spectrometry, atomic force microscopy and X-ray diffraction. Deuterium implantation was performed using a deuterium plasma jet and its retention in nitrogen containing tungsten films was investigated using thermal desorption spectrometry. Deuterium retention and release behaviour strongly depend on the nitrogen content in the coatings and the films microstructure. All nitride coatings have a polycrystalline structure and retain a lower deuterium level than the pure tungsten sample. Nitrogen content in the films acts as a diffusion barrier for deuterium and leads to a higher desorption temperature, therefore to a higher binding energy. [ABSTRACT FROM AUTHOR]

Deuterium plasma has been produced in a cylindrical inertial electrostatic confinement fusion device using hot and cold cathode discharges and the plasma parameters are determined by employing an electrostatic probe. The plasma temperature and density are estimated at optimum experimental conditions and it is noted that the plasma temperature is 3 eV in the case of hot cathode discharge whereas 10 eV in the case of the cold cathode discharge. The plasma density as determined is two orders more in the case of the hot cathode discharge than the other. The probe is also used to observe the ion oscillation in the negative potential well that is formed in between the cathode grid and chamber (anode). The observation of spontaneous oscillation along with the harmonics has been reported. [ABSTRACT FROM AUTHOR]

Micro nuclear reaction analysis (micro-NRA) exploiting the nuclear reaction D( 3 He,p) 4 He was used for post-mortem analyses of special marker samples, exposed to deuterium plasma inside ASDEX Upgrade (AUG) tokamak and to the deuterium plasma jet in the Pilot-PSI linear plasma gun. Lateral concentration profiles of deuterium and erosion/deposition profiles of the marker materials were obtained by a combination of micro-NRA and particle induced X-ray emission by 3 He beam (3HIXE). In the case of AUG samples, where 25 nm thick W marker layers had been deposited on unpolished and polished graphite substrates, the effect of surface roughness on local erosion and deposition was also investigated. The lateral distribution of W concentration showed that erosion is much more distinct in the case of polished samples and the resulting surface shows a “leopard” skin pattern of W accumulated on carbon aggregates left on the surface from polishing. The Pilot-PSI samples indicated preferential accumulation of deuterium a few mm off from the centre of the region affected by the plasma beam. This is connected with the largest surface modifications while the thick deposited layers at the centre do not favour deuterium retention per se. The results were cross correlated with those obtained using laser-induced breakdown spectroscopy (LIBS). With its quantitative abilities, micro-NRA provided essential calibration data for in situ LIBS operation, as well as for complementary post mortem Secondary Ion Mass Spectroscopy (SIMS). [ABSTRACT FROM AUTHOR]

Experiments on deuterium retention on liquid lithium confined in a capillary structure followed by ex-situ thermal desorption spectrometry (TDS) at high plasma fluxes (∼10 23 m 2 s −1 ) and high temperatures (440 °C and 580 °C) have been performed. Deuterium plasmas were generated at the PILOT-PSI linear plasma device and the targets were a 30 mm diameter stainless steel disc, 5 mm thick, covered with a porous mesh and filled with lithium. The settings (current) of the plasma source were varied in order to get different sample surface temperatures during irradiation. The targets were kept at floating potential during the exposure. Hydrogen and Li emission signals were monitored during the plasma exposure and TDS analysis was made afterwards in a separated system. Decreased retention at high exposure temperatures was deduced from the analysis of the hydrogen emission signals. Nevertheless, the results from TDS signal analysis were not conclusive. [ABSTRACT FROM AUTHOR]

This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10 −13 cm −2 s −1 . The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT) , and above 623 K − K = 10 9 exp(-144/RT) . The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. [ABSTRACT FROM AUTHOR]

The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary. [ABSTRACT FROM AUTHOR]

tungsten samples were heat-treated to achieve partial recrystallization and exposed to high ion flux deuterium plasma at different temperatures and fluences. Continuous stiffness nanoindentation measurements of near-surface hardness were performed in the grains of specific annealing states and of specific crystallographic orientation, determined by electron backscatter diffraction (EBSD); indentation pile-up was investigated using surface profilometry. Bulk hardness of unexposed tungsten does not strongly depend on grain orientation, but depends on the annealing state of the grain, with values between ∼4.3 GPa for recrystallized grains and ∼5.5 for non-recrystallized ones. Grains with <111> surface normal orientation feature the least pile-up, while grains with <001> orientation the most; pile-up also depends on the annealing state, being generally lower in recrystallized grains. Plasma exposure leads to the increase of hardness, most significantly near the surface. The width of plasma-affected zone increases with the increase of exposure temperature and fluence, as well in recrystallized grains, correlating with the increase of diffusion depth. Plasma exposure does not lead to the emergence of orientation-dependence of hardness. Both indentation pile-up and near-surface indentation pop-ins are generally suppressed by plasma exposure. [ABSTRACT FROM AUTHOR]

At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium–tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle–energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons. [ABSTRACT FROM AUTHOR]

The deuterium plasma experiment using Large Helical Device (LHD) has been conducted since 2017. A small amount of tritium was produced in the vacuum vessel and a part of tritium remained in the plasma-facing materials. After the deuterium plasma experimental campaign, the vacuum vessel was opened for the LHD maintenance activity. Then, the plasma-facing material was exposed to room air and remained tritium was released at ambient temperature. The tritium release rate increased rapidly immediately after the vacuum vessel was opened to the atmosphere, but it turned to decrease soon after. After several weeks of opening the vacuum vessel, the tritium release rate became low and almost constant. The total tritium decontamination rate, defined as the ratio of tritium released by air exposure to tritium produced during the plasma experimental campaign, was achieved at several percent. The air exposure operation is expected to have a decontamination effect due to isotope exchange reactions on the surface of the plasma-facing material. [ABSTRACT FROM AUTHOR]

• The formation of surface blisters and deuterium retention in tungsten preliminarily irradiated by argon ions and then exposed to deuterium plasma were studied in this work. • An increase in depth and an increase in the level of argon-induced damage lead to a decrease in the area density of blisters and an increase in deuterium retention. • The reduced surface blisters are mainly in diameter below 10 µm, suggesting that deuterium aggregation at the grain boundaries in the damaged region was suppressed due to the increase in the number of lattice defects created by argon ion pre-irradiations. • Thermal desorption spectra in similar shapes implicate that the type of irradiation-induced defects was rarely affected by the ion energy as well as fluence used in this work. Heavy ions have been widely used to simulate the effect of neutron irradiation-induced crystal defects on the plasma-wall interaction in the future fusion device. In this work, energetic argon ions were utilized to damage tungsten with the different pre-damage distributions. The pre-damage distribution is controlled by tuning ion energy and fluence. After exposure to deuterium plasma, the pre-damaged tungsten targets were analyzed to explore the impact of pre-damage distribution on deuterium-induced surface blistering and retention. Surface observation results reveal that increasing pre-damage depth and peak dpa significantly reduce the surface blister's area density, especially in diameters below ten micrometers. Meanwhile, thermal desorption spectra indicate that the given change in the pre-damage distribution barely change the type of irradiation-induced defects. The role of pre-damage increases in deuterium trapping and aggregation is discussed based on the experimental results. [ABSTRACT FROM AUTHOR]

Diamond's intrinsic hardness, excellent thermal conductivity and low atomic number make it a highly promising candidate as a plasma facing material. However, as with the previously used graphite, concerns over tritium retention and resultant chemical etching have so far limited research interest in the use of synthetic diamond. In order to study tritium retention, the DELPHI facility at the Culham Centre for Fusion Energy was used to expose polycrystalline diamond samples to a deuterium plasma. Deuterium ions were accelerated to an energy of 0.2 keV to 1 keV for a 5 h exposure time, achieving a fluence of approximately 5.5 × 1021 D m−2. Exposed samples were analysed using Thermal Desorption Spectroscopy. Increasing implantation energy resulted in additional D 2 release peaks observed in the 800–1100 K temperature range that were not seen at lower energies. These peaks were interpreted as an additional bonding mechanism, a likely candidate for which is inter-grain deuterium. Experimental work was complemented with molecular dynamics simulations on the University of Bristol's high performance computer—Blue Crystal Phase 4. In these simulations, both varying implantation energy and the presence of grain boundaries were explored. A two-step etching mechanism was observed, in which the surface initially swelled before carbon removal. No significant differences could be observed on the inclusion of a grain boundary at the energies tested. • Synthetic diamond exposed to 200–1000 eV deuterium ions. • Deuterium retention determined with thermal desorption spectroscopy. • Inter-grain penetration observed for ion energies >400 eV. • Two-step etching mechanism observed with molecular dynamics simulations. • Retention values comparable to other metallic plasma facing material candidates. [ABSTRACT FROM AUTHOR]

Tungsten is a primary candidate for plasma facing materials for future fusion devices. An important safety concern in the design of plasma facing components is the retention of hydrogen isotopes. Available experimental data is vast and scattered, and a consistent physical model of retention of hydrogen isotopes in tungsten is still missing. In this work we propose a model of non-equilibrium hydrogen isotopes trapping under fusion relevant plasma exposure conditions. The model is coupled to a diffusion-trapping simulation tool and is used to interpret recent experiments involving high plasma flux exposures. From the computational analysis performed, it is concluded that high flux high temperature exposures (T = 1000 K, flux = 10 24 D/m 2 /s and fluence of 10 26 D/m 2 ) result in generation of sub-surface damage and bulk diffusion, so that the retention is driven by both sub-surface plasma-induced defects (bubbles) and trapping at natural defects. On the basis of the non-equilibrium trapping model we have estimated the amount of H stored in the sub-surface region to be ∼10 −5 at −1 , while the bulk retention is about 4 × 10 −7 at −1 , calculated by assuming the sub-surface layer thickness of about 10 μm and adjusting the trap concentration to comply with the experimental results for the integral retention. [ABSTRACT FROM AUTHOR]