• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1706-1713
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• 2020

Functionally gradient material (FGM) is promisingly effective in mitigating the thermal stress between plasma facing materials (PFM) and structural materials. However, the corresponding research with respect to W/V FGM has not been reported yet. In this work, we firstly report the successful fabrication of W/V FGM by a combined technology of mechanical alloying (MA) and spark plasma sintering (SPS). The microhardness and microstructure of the consolidated sample were both investigated. W/V stacks show significantly enhanced microhardness (>100%) compared with pure W plate, which is beneficial to the integral strength of the hybrid structure. Furthermore, we clarify that the different ductility of W and V should be carefully considered, otherwise W/V powder might aggregate and lead to the formation of compositional segregation, and simultaneously unmask the impact of V proportion on the distribution of second phase in W-V binary alloy system. This work provides an innovative approach for obtaining W-V connections with much better performance.

• Korean Journal of Metals and Materials
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• v.48 no.3
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• pp.268-275
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• 2010

A new approach is described for preparing AlN thin films containing various amounts of Co content by using a two-facing targets type sputtering (TFTS) system. The deposited films were annealed isothermally at different temperatures and their microstructure, magnetic properties and resistivity were investigated. A small saturation magnetization ($4{\pi}Ms=0.52{\sim}0.85kG$) was observed irrespective of Co content in the asdeposited films. It was found that annealing conditions can control physical properties as well as the microstructure of the films. A high saturation magnetization (3.7 kG) and resistivity of $2200{\mu}{\Omega}-cm$ was obtained for AlN films containing 25 at.% Co.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2141-2152
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• 2024

Tungsten is the most promising plasma facing material for fusion reactors. Rolled W-Y₂(Zr)O₃ bulk material has been successfully produced in this study for future fusion engineering applications. The introduction of Zr is conducive to the refinement of the second phase particles. Nano-sized Y-Zr-O particles are observed in the powder and bulk samples. Related results show that the Y-Zr-O particle dispersion distribution improves the heat load resistance of W-Y₂(Zr)O₃ composite material. For four-point bend experiments in the same sampling direction, the DBTT of W-Y₂(Zr)O₃ composite materials is lower compared to the pure tungsten. For the same material, the DBTT of the material was selected for testing along the RD direction is lower compared to the material was selected for testing along the TD direction. Findings of this study provide suggestions for the subsequent industrial preparation of nanoscale particle-doped tungsten materials.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.827-834
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• 2020

Damage of tungsten due to helium ions of a PF device was studied. The tungsten was analyzed by SEM and AFM after irradiation. SEM revealed fine bubbles of helium atoms with diameters of a few nanometers, which join and form larger bubbles and blisters on the surface of tungsten. This observation confirmed the results of molecular dynamics simulation. SEM analysis after etching of the irradiated surface indicated cavities with depth range of 35-85 nm. The average fluence of helium ion of the PF device was calculated about 5.2 × 10¹⁵ cm^-2 per shot, using Lee code. Energy spectrum of helium ions was estimated using a Thomson parabola spectrometer as a function of dN/dE ∝ E^-2.8 in the energy range of 10-200 keV. The characteristics of helium ion beam was imported to SRIM code. SRIM revealed that the maximum DPA and maximum helium concentration occur in the depth range of 20-50 nm. SRIM also showed that at depth of 30 nm, all of the tungsten atoms are displaced after 20 shots, while at depth of higher than 85 nm the destruction is insignificant. There is a close match between SRIM results and the measured depths of cavities in SEM images of tungsten after etching.

• Smart Structures and Systems
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• v.24 no.6
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• pp.759-768
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• 2019

In nuclear fusion reactors, the key structural component (i.e., the plasma-facing component) undergoes high heat flux cyclic loading. To ensure the safety of fusion reactors, an experimental study on the temperature-induced creep of stainless steel under heat flux cyclic loading was performed in the present work. The strains were measured using a stereo digital image correlation technique (3D-DIC). The influence of the heat haze was eliminated, owing to the use of a vacuum environment. The specimen underwent heat flux cycles ($500^{\circ}C-1000^{\circ}C$) with different mechanical preloads (0 kN, 10 kN, 30 kN, and 50 kN). The results revealed that, for a relatively large preload (for example, 50 kN), a single temperature cycle can induce a residual strain of up to $15000{\mu}{\varepsilon}$.

• Journal of the Korean Society for Heat Treatment
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• v.20 no.5
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• pp.231-236
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• 2007

High velocity oxy-fuel (HVOF) spraying was used to coat Ti(Al,O)/$Al_2O_3$ powder onto the SS41 steel plate. Macrostructure of the coated specimen has been investigated by scanning electron micrograph (SEM). High temperature oxidation behavior of the coated specimen and SS41 steel have been studied. From the results of SEM observation, Ti(Al,O)/$Al_2O_3$ powder was coated well onto the substrate SS41 steel. Porosity onto the coated layer was only 0.38%. The oxidation results showed that Ti(Al,O)/$Al_2O_3$ powder coated SS41 steel have improved little oxidation resistance at $900^{\circ}C$ in air, but improved remarkably oxidation resistance at $800^{\circ}C $ in air compare to the substrate SS41 steel.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1490-1500
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• 2024

Helium-induced defect nucleation and accumulation in polycrystalline W and W0.5 wt%ZrC (W0.5ZrC) were studied in-situ using the transmission electron microscopy (TEM) combined with 40 keV He⁺ irradiation at 800 and 1000℃ at the maximum damage level of 1 dpa. Radiation-induced dislocation loops were not observed in the current study. W0.5ZrC was found to be less susceptible to irradiation damage in terms of helium bubble formation and growth, especially at lower temperature (800 ℃) when vacancies were less mobile. The ZrC particles present in the W matrix pin the forming helium bubbles via interaction between C atom and neighbouring W atom at vacancies. This reduces the capability of helium to trap a vacancy which is required to form the bubble core and, as a consequence, delays, the bubble nucleation. At 1000 ℃, significant bubble growth occurred in both materials and all the present bubbles transitioned from spherical to faceted shape, whereas at 800 ℃, the faceted helium bubble population was dominated in W.

• Journal of the Semiconductor & Display Technology
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• v.5 no.1 s.14
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• pp.13-16
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• 2006

AIN thin films were prepared on amorphous glass and $SiO_2(1{\mu}m)/Si(100)$ substrate by the facing targets sputtering (FTS) apparatus, which can provide high density plasma, a high deposition rate at a low working gas pressure. The AIN thin films were deposited at a different nitrogen gas flow rate ($1.0{\sim}0.3$) and other sputtering parameters were fixed such as sputtering power of 200w, working pressures of 1mTorr and AIN thin film thickness of 800 nm, respectively. The thickness and crystallographic characteristics of AIN thin films as a function of $N_2$ gas flow rate $[N_2/(N_2+Ar)]$ were measured by $\alpha$-step and an X-ray diffraction (XRD) instrument. And the c-axis preferred orientations were evaluated by rocking curve. In the results, we could prepared the AIN thin film with c-axis preferred orientation of about $5^{\circ}$ on substrate temperature R.T. at nitrogen gas flow rate 0.7.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4491-4498
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• 2022

The thermal stress between W plasma-facing material (PFM) and Cu heat sink in fusion reactors can be significantly reduced by using a W-Cu functionally graded material (W-Cu FGM) interlayer. However, there is still considerable stress at the joining interface between W and W-Cu FGM in the W/W-Cu FGM/Cu portions. In this work, we fabricate W skeletons with continuous gradients in porosity by a modified sedimentation method. Sintering densification behavior and pore characteristics of the sedimented W skeletons at different sintering temperatures were investigated. After Cu infiltration, the final W-Cu FGM was obtained. The results indicate that the pore size and porosity in the W skeleton decrease gradually with the increase of sintering temperature, but the increase of skeleton sintering temperature does not reduce the gradient range of composition distribution of the final prepared W-Cu FGM. And W-Cu FGM with composition distribution from pure W to W-20.5wt.% Cu layer across the section was successfully obtained. The thickness of the pure W layer is about one-fifth of the whole sample thickness. In addition, the prepared W-Cu FGM has a relative density of 94.5 % and thermal conductivity of 185 W/(m·K). The W-Cu FGM prepared in this work may provide a good solution to alleviate the thermal stress between W PFM and Cu heat sink in the fusion reactors.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2139-2146
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• 2023

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 ℃ and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ~268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m² for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.