Reduced Activation Ferritic Steel Research Articles

Microstructural Evolution and High‐Temperature Tensile Properties of 15Cr‐Reduced Activation Ferritic Steel Processed by Hot Powder Forging of Mechanically Alloyed Powders

Reduced activation ferritic steels are being explored as possible cladding tube materials for nuclear reactors because of their low activation and excellent irradiation resistance. In the current investigation, reduced activation ferritic steel (Fe–15Cr–2W) is processed by mechanical alloying of elemental powders followed by hot powder forging. Mechanical alloying is carried out in a Simoloyer attritor mill (Zoz GmbH), after which the powders are placed in a mild steel can and forged at 1200 °C in H2 atmosphere. X‐ray diffraction and transmission electron microscopy (TEM) investigation reveal that 10 h of mechanical alloying is required to achieve complete dissolution of Cr and W in the Fe matrix powder. The relative density and hardness distribution of the forged slab is evaluated in longitudinal as well as transverse direction to optimize the powder forging operation. Electron backscatter diffraction analysis showed dynamic recrystallization to take place during the course of hot powder forging. Tensile tests are performed at room temperature as well as at elevated temperatures (600 and 700 °C). The yield strength and ultimate tensile strength at room temperature as well as at elevated temperatures are found to be higher than those reported in literature for reduced activation ferritic steels consolidated by other techniques.

Tensile and creep properties of small specimens of reduced-activation ferritic steel F82H, and the correlation to standard specimen data

Background A reduced-activation ferritic/martensitic (RAFM) steel, F82H steel, is the primary candidate structural material for fusion reactor blanket. Small specimen test technique is essential to develop the blanket materials using limited irradiation volume in high flux neutron field. An international collaboration activity “Towards the Standardization of Small Specimen Test Techniques for Fusion Applications” has been initiated under the framework of the International Atomic Energy Agency Coordinated Research Project for Phase I from 2017 to 2021, and Phase II from 2022 to 2026. The present paper reports the preliminary results on tensile and creep tests as a summary of the above Phase I activity. Methods Tensile and creep tests were conducted at 550 and 650°C, using flat-plate SSJ type small specimens with various gauge thickness ranged from 0.14 to 1.2 mm, while gauge length and width are 5 and 1.2 mm, respectively. In addition, round bar type standard specimens with a gauge geometry of 6 mm in diameter and 30 mm in length was also tested for comparison. Results Tensile yield stress, ultimate tensile strength and uniform elongation were independent of the gauge thickness of SSJ specimens, and agreed with the data from the standard size specimens. On the other hand, total elongation was decreased with decreasing the thickness. In creep tests, rupture time was decreased with decreasing the gauge thickness of SSJ specimens. Standard size specimens exhibited shorter rupture time than the SSJ specimens. Conclusions The SSJ type specimens provided similar tensile parameters to those from the standard specimen, except total elongation. Creep rupture time of the SSJ specimens were different from the standard specimen, and decreased with decreasing the gauge thickness.

Effects of thermal aging at 873 K on the impact properties of an ODS ferritic steel

This investigation presents the effects of long-term thermal aging on the impact properties of an ODS reduced activation ferritic steel. The ODS ferritic steel, with composition Fe–14Cr–2W–0.4Ti–0.3Y2O3 (wt. %), was manufactured by mechanical alloying, compacted by hot isostatic pressing and hot cross rolled. A 1273 K annealing treatment was applied, followed by thermal aging at 873 K for 2000 h in Ar, to evaluate the thermal stability of the alloy. Charpy impact testing was performed from 77 to 473 K to investigate the mechanical response of the material under dynamic loads. The ductile-to-brittle transition temperature (DBTT) showed a slight increase with values of 255 ± 12 K before aging, and 270 ± 20 K after aging. The aging treatment softened the steep character of the Charpy impact curves inducing a wider transition regime with mixed brittle and ductile fracture mechanisms. This behavior was confirmed by the analysis of Force vs. Displacement Charpy curves and fractographic observations. Hot cross rolling induced a crack divider geometry and secondary cracks formation by delamination. Ductile fracture was less evident after aging due to the redistribution and transformation of Cr-W-rich precipitates along grain boundaries. Electron Backscattered Diffraction highlighted the features of transgranular brittle cracks at temperatures below the DBTT and showed misoriented and deformed regions near the fracture surface at temperatures above the DBTT.

Effect of tantalum addition on the mechanical properties of tungsten/tantalum composite thin films

Tungsten (W) when used as a plasma-facing material (PFM) in a tokamak will face harsh neutron radiation and high-temperature fluctuations. Due to its ductile-to-brittle transition temperature (DBTT), W behaves as a brittle material in lower temperatures and thus helps in propagating cracks easily resulting in dust formation and failure of material. In this study, an attempt has been made to improve the ductility, hardness, and toughness of W thin films by adding tantalum (Ta) to it, resulting in composite thin films. W/Ta composite thin films deposited on reduced-activation ferritic martensitic steel (RAFM) steel by RF magnetron sputtering method are compared with W thin films deposited on the same substrate. Taguchi orthogonal array is used to get the best properties for the composite thin films and nanoindentation technique is used to evaluate the mechanical properties of developed thin films. With the least hardness of 7.19 Gpa, composite film hardness is 452% lesser than the maximum hardness of (39.54 Gpa) W films. The reduction in hardness for composite thin films has not depleted the toughness as the Ef/Es ratio is higher than cracking threshold of 1.3. The maximum plastic energy absorbed for yield by composite films during nanoindentation is (1710 mN nm) 14% higher than W thin film (1500 mN nm), indicating the composite film is a ductile and tougher than the W film.

Influence of surface temperature, ion impact energy, and bulk tungsten content on the sputtering of steels: In situ observations from plasma exposure in PSI-2

• PSI-2 deuterium plasma exposures of 3 RAFM steels. • In-situ measurement of relative sputtering yields by optical emission spectroscopy. • In-situ measurement of the evolution of surface morphology by IR camera. • Post-mortem measurement of tungsten enrichment by RBS, SEM and EDX. • Independent measurements allow to disentangle the impact on the erosion of Fe. This study presents results from PSI-2 deuterium plasma exposures of reduced-activation ferritic martensitic steels with different tungsten content at a peak plasma flux of 9 · 10 20 D/m 2 s and 420–920 K sample temperature. Via optical emission spectroscopy the relative sputtering yield of Fe and Cr, and via an infrared camera the evolution of surface morphology is studied in-situ in 2D. After the exposure, RBS, SEM and EDX reveal tungsten enrichment and the formation of surface structures. Due to the independent measurement of surface morphology and tungsten enrichment, it is possible to disentangle the impact of both effects on the erosion of Fe. As part of this study, we investigate the effects of tungsten content, sample temperature and ion impact energy. The results show a reduction of the sputtering yield, depending on the exposure conditions, up to a factor of 2 at 130 eV ion impact energy and 720 K sample temperature.

Benchmarking a 9Cr-2WVTa Reduced Activation Ferritic Martensitic Steel Fabricated via Additive Manufacturing

Reduced activation ferritic (RAF) martensitic steels are promising candidates for the first wall of fusion reactors. However, current manufacturing capabilities call for these components to be made by welding wrought plates. This limits design freedom and necessitates the use of post-weld heat treatments (PWHT) in accordance with the boiler and pressure vessel code. Additive manufacturing (AM) can offer a unique solution to solve this challenge by leveraging the layer-wise deposition strategy to come up with temper bead deposition techniques to eliminate post-processing heat treatments (PPHT). However, it is necessary to benchmark the properties of RAF steels fabricated by AM with their wrought counterparts to identify the process-structure-property correlation, which is the goal of this study. The study demonstrates that while tensile properties at room temperature and high temperatures are satisfactory, the as fabricated and samples after PPHT have significant heterogeneity in tensile elongation. This has been attributed to the presence of discontinuities in the build. The as-fabricated samples have an average tensile strength of 1190 + 12 MPa and an average elongation of 15 + 5% at room temperature and 658 ± 20 MPa ultimate tensile strength (UTS) and 14 ± 7% at 600 °C. After the post-weld heat treatment, mechanical properties decrease to around 600–650 MPa and an elongation between 20–25% at room temperature to 300 MPa UTS and 25–28% elongation at 600 °C. The characterization of microstructures at various length scales demonstrates that the as-fabricated structure has a significant fraction of delta ferrite in a lath martensitic matrix. No precipitates could be identified in the as-fabricated structure. PPHT led to a decrease in the area fraction of delta ferrite and precipitation of M23C6 and MX. Detailed characterization clearly demonstrates that the lack of precipitates in the as-fabricated structure could be due to the slow tempering response of the alloy. Finally, the needs to develop new alloys to achieve the objectives stated above are articulated.

Powder Particle Size Effects on Microstructure and Mechanical Properties of Mechanically Alloyed ODS Ferritic Steels

Reduced activation ferritic (RAF) steels are expected to be widely used in challenging nuclear industrial applications under severe thermo-mechanical regimes and intense neutron loads. Therefore, actual research panorama is facing the strengthening strategies necessary to maximize both performance and endurance under these conditions. Oxide dispersion strengthened (ODS) RAF steels are leader candidates as structural materials in fusion energy reactors thanks to the reinforcement obtained with a fine dispersion of nanosized oxides in their matrix. In this study, the influence of the initial powder particle size and the selected processing route on the final material has been investigated. Two RAF ODS steels coming from atomized pre-alloyed powders with nominal particle powder sizes of 70 and 30 µm and composition Fe-14Cr-2W-0.4Ti-0.3Y2O3 (wt. %) were manufactured by mechanical alloying. Alloyed powders were compacted by hot isostatic pressing, hot crossed rolled, and annealed at 1273 K. Initial powder particle size differences minimize after milling. Both steels present an almost completely recrystallized material and similar grain sizes. The same type and distributions of secondary phases, Cr-W-rich, Ti-rich, and Y-Ti oxide nanoparticles, have been also characterized by transmission electron microscopy (TEM) in both alloy samples. The strengthening effect has been confirmed by tensile and Charpy impact tests. The two alloys present similar strength values with slightly better ductile brittle transition temperature (DBTT) and ductility for the steel produced with the smaller powder size.

Effect of thermomechanical treatment on the microstructure and mechanical properties of Si-containing oxide dispersion strengthened reduced activation ferritic steel synthesized via mechanical alloying and spark plasma sintering

Abstract The present work describes the effect of thermo-mechanical treatment on the microstructure, and its subsequent effect on the mechanical properties of Si-containing oxide dispersion strengthened (ODS) reduced activation ferritic (RAF) steel. ODS-RAF steel powder mixture with a nominal composition of Fe-14Cr-2W-0.3Ti-1Si-0.3Y2O3 was mechanically alloyed up to 50 h and subsequently consolidated via spark plasma sintering (SPS) at 900 °C followed by thermo-mechanical treatment (TMT) at 850 °C. Microstructural study revealed that the sintered sample consisted of ultrafine grains (0.35 µm) along with nano-size particles (Cr2TiO4, SiO2, and Y2Ti2O7) distributed in the matrix. This sintered sample exhibited excellent mechanical properties such as Vickers hardness (600 VHN), compressive strength (1795 MPa), and total elongation (21%), which is ascribed to the presence of ultrafine grains and nano-size particles. The thermo-mechanically treated sample exhibited an increase in the density by 2.7% and a reduction in the grain size by 23%. Consequently, the Vickers microhardness, compressive strength, and failure elongation of the sintered sample improved by 13%, 37%, and 19%, respectively, upon thermo-mechanical treatment.

On the axisymmetric stability of tokamaks with ferromagnetic walls

Reduced activation ferritic steels are an attractive option for use in large structural components surrounding tokamak plasmas in future fusion power plants, but their ferromagnetic response to the confining magnetic fields must be properly understood. Simultaneously, the advantages of operating at high plasma elongation push tokamak designs toward scenarios that are more vulnerable to vertical displacement events. Passive conducting structures in present tokamaks slow these instabilities such that they may be feedback controlled, but the efficacy of this process is likely to be eroded by ferromagnetic effects. We approach two related analytical models—in cylindrical and spherical geometries—which qualitatively and quantitatively assess the impact of a ferritic steel wall on the vertical instability growth rate for a plasma of certain elongation. Distinct limits for magnetically thick and thin walls give key physical insight, but the dependence on magnetic permeability and wall geometry is, in general, quite complex. Equilibrium considerations, particularly with respect to radial force balance, are also encountered.

Thermomechanical Solid Breeder Multiple-Effects Experiment (TESOMEX) with Volumetric Heating and FEM Benchmark Analysis

Available experimental efforts on ceramic pebble beds and their associated constitutive equations are necessarily derived from single-effect tests where one parameter is varied and its effects are isolated and studied separately (e.g., using constant temperatures and externally applied loads). These experiments are incapable of reproducing the true multiple/synergistic effects of the physics that occur in real blankets, and the phenomena arising from the interactions of single effects are yet to be discovered. It is unclear whether the combined effect of plasticity and creep under reactor-relevant loading conditions will either enable the altered pebble bed packing configuration to reach an acceptable self-regulating temperature state or significantly deteriorate its heat transfer efficiency and subsequent tritium release. Therefore, studying the isolated thermal and mechanical effects is not sufficient to predict pebble bed behavior. It is the coupling and interdependence between the dynamic thermal and mechanical fields, as well as the synergistic effects between the various modes of deformation, that are key to fully understanding and predicting pebble bed behavior in a realistic fusion environment. However, previous mock-up experimental campaigns thus far have suffered from critical shortfalls that have severely hamstrung their scientific impact. The lack of experimental data that incorporate multiple-effects interactions in addition to the complexity of building a full-scale breeder unit mock-up triggered the need for this experimental effort. The body of work presented in this paper serves to (1) establish and validate the practicality of various volumetric heating simulation techniques for representative thermomechanical study, (2) recreate a prototypical breeder unit’s thermal-hydraulic behavior using a scaled-down reduced-activation ferritic steel box with optimized manifold design, (3) evaluate the thermomechanical properties of the pebble bed using a novel nonintrusive in situ tactile-pressure-sensing technology capable of generating real-time contact pressure maps that reveal spatial and temporal stress evolution, and (4) develop and benchmark a thermomechanical finite element method code that is able to predict the pebble bed’s thermomechanical evolution under the effects of creep and thermal cycling. The results of this study not only present novel experimental techniques and data that enhance our understanding of synergistic thermomechanical interactions and effects, but they also provide valuable data to serve as a basis for validation of the most recent pebble bed numerical models. Finally, it is worth mentioning that this work is part of a compendium around the Thermomechanical Solid Breeder Multiple Effects Experiment experimental campaign, known as TESOMEX. While this paper primarily focuses on novel heating and instrumentation techniques along with their opportunities and limitations, the other two papers shed more light on the prototypical thermomechanical evolution, pebble sintering, and possible modes of failure.

Deuterium concentration depth profiling in sputter-deposited tungsten coated F82H using secondary ion mass spectrometry

Abstract Deuterium plasma-driven permeation (PDP) through reduced activation ferritic steel alloy F82H with and without sputter-deposited tungsten (SP-W) coatings has been conducted using a laboratory-scale steady-state plasma facility. SP-W coatings have been found to enhance PDP flux with a much slow breakthrough to reach steady-state. Static deuterium retention has been evaluated with thermal desorption spectroscopy. Enhanced deuterium retention in SP-W coated F82H has been observed, compared with that in bare F82H. Secondary ion mass spectrometry (SIMS) has been carried out to obtain the depth profile of deuterium concentration crossing the W/F82H interface. SIMS data indicate the deuterium concentration is much higher in W coatings than that in F82H substrate, exhibiting an “uphill diffusion” profile.

Erosion and fuel retentions of various reduced-activation ferritic martensitic steel grades exposed to deuterium plasma

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.

Oxide layer formation in reduced activation ferritic steel F82H under DEMO reactor blanket condition

Tritium permeation through structure materials in fusion blanket systems is a critical issue from the perspectives of fuel loss and radiological hazard. In the previous studies, detailed hydrogen isotope permeation behaviors in reduced activation ferritic/martensitic steels have been investigated; however, oxidation of the steel surface is expected under an actual DEMO reactor condition, and then the tritium permeation behavior will be changed. In this study, deuterium permeation through the steels heat-treated under simulated environment conditions has been investigated for more precise predictions of tritium loss at DEMO reactor blankets. Reduced activation ferritic/martensitic steel F82H substrates were heat-treated in helium gas flow containing 1 vol% hydrogen at 300, 400 and 500 °C for 100 and 200 h to simulate a solid breeder DEMO reactor blanket condition. After surface observation and analysis for the heat-treated samples, gas-driven deuterium permeation measurements were performed. An iron oxide layer was formed on the sample surface, and the thickness of the layer was 50 nm‒12 μm. The oxide layer on the sample surface heat-treated at 500 °C for 100 h decreased deuterium permeation by a factor of 5. After the permeation tests, dissipation of the oxide layers was confirmed.

The effect of γ-ray irradiation on deuterium permeation through reduced activation ferritic steel and erbium oxide coating

Deuterium permeation through a fusion-relevant ferritic steel F82H with and without erbium oxide coatings under γ-ray irradiation has been investigated in a temperature range from 300 to 700 °C. The deuterium permeation flux through F82H sample increased by γ-ray irradiation at lower temperatures below 450 °C. The irradiation effect increased with dose rate, and the percentage of the permeation flux gain might be several percent under the dose rate of a few Gy s‒1. Temperature of the F82H sample surface rose by about 0.5 °C depending on the dose rate, and so the γ-ray irradiation effect is mainly attributed to γ-heating. On the other hand, at higher temperature above 500 °C, no appreciable change of the deuterium permeation was observed. Similarly, the deuterium permeation flux through erbium oxide coated samples increased under γ-ray irradiation at lower temperatures (350‒450 °C), but no appreciable change of permeation flux through coatings was observed at higher temperatures (600‒700 °C). The coating surface temperature increased at lower sample temperatures by γ-heating.

Micro-tensile testing of reduced-activation ferritic steel F82H irradiated with Fe and He ions

Micro-tensile tests were conducted on reduced-activation ferritic steel F82H specimens after irradiation with Fe and He ions. The displacement damage levels were 50 and 200 dpa, and helium concentrations were 0, 2000 and 5000 appm. The specimens had dimensions of 8 × 1 × 1 µm for the gage section and they were tension tested at room temperature in a vacuum. The yield strengths of the unirradiated specimens were close to the literature value reported for a large, unirradiated specimen. The increases in yield strength and ultimate tensile strength due to Fe ion irradiation were clearly observed. The loss of work hardening was confirmed for the 200 dpa specimens (0 or 2000 appmHe). Higher yield strength and ultimate tensile strength were confirmed for the 50 dpa/5000 appmHe specimen compared to the no helium specimen (50 dpa/0 appmHe). However, no apparent helium effects were confirmed for the 200 dpa/2000 appmHe specimen by the micro-tensile test at room temperature.

HiperFer, a reduced activation ferritic steel tested for nuclear fusion applications

Materials are the most urgent issue in nuclear fusion research. Besides tungsten, steels are considered for unifying functional and structural materials due to their cost and mechanical advantages over tungsten. However, the fusion neutrons impose a strong constraint on the ingredients of the steel in order to avoid long lasting activation, while the material has to pertain sputtering resistance, low hydrogen retention, and long-term mechanical stability. In this proof-of-principle, we demonstrate the interesting properties of the new material HiperFer (High performance Ferrite) as a material suitable for fusion applications.The investigation covers neutron activation modelled by FISPACT-II, plasma sputtering and deuterium retention experiments in PSI-2, thermo-mechanical properties and component modelling. The material was found to feature a low nuclear inventory. Its sputtering yield reduces due to preferential sputtering by a factor 4 over the PSI-2 D2 plasma exposure with possible reductions of up to 70 indicated by SD.Trim.SP5 modelling. The exposure temperature shows a strong influence on this reduction due to metal diffusion, affecting layers of 1 µm in PSI-2 at 1150 K exposure for 4 h. Deuterium retention in the ppm range was found under all conditions, together with ∼10 ppm C and N solubility of the ferritic material. The creep and cyclic fatigue resistance exceed the values of Eu-97 steel. As an all HiperFer component, heat loads in the order of 1.5 MW/m² could be tolerated using water-cooled monoblocks. In conclusion, the material solves several contradictions present with alternative reduced-activation steels, but its applications temperatures >820 K also introduce new engineering challenges.

Characterisation of ODS Fe-14Cr-2W-0.3Ti before and after high temperature triple and low temperature single ion irradiations

Oxide dispersion strengthened reduced activation ferritic steels are being considered as structural materials for future fusion reactors and Gen IV fission reactors. In this work, the stability of the nanometric oxides after being irradiated at different conditions and temperatures has been investigated. A Fe-14Cr-2W-0.3Ti-0.3Y2O3 alloy was simultaneously triple-beam irradiated at 600°C with Fe, He and H ions. The same alloy was irradiated with Fe ions at −80°C. The microstructure of this alloy was investigated using Transmission Electron Microscopy, Atom Probe Tomography and nanoindentation before and after the irradiations. There were no significant changes in the morphology and chemical composition of the nanoparticles after the irradiations, although the size of the smaller ones tends to decrease for both irradiation conditions. Small irradiation-induced bubbles were present after the triple simultaneous irradiation at high temperature. Nanoindentation also shows no significant differences with a slight increase in hardness for the sample irradiated at low temperature, while softening appears in the high temperature irradiation.

Effect of hot cross rolling on the microstructure and mechanical properties of an Fe-14Cr ODS ferritic steel

Oxide dispersion strengthened (ODS) reduced activation ferritic steels are leading candidates to become part of the structure of the first wall/blanket of future fusion reactors. Their major drawback is their poor toughness, which must be improved. In this work, an ODS ferritic steel having nominal composition Fe–14Cr–2W–0.2Ti–0.55 Fe2Y (wt%), produced by mechanical alloying of prealloyed powders with Fe2Y intermetallic particles and consolidated by hot isostatic pressing, was subjected to either thermal treatments at high temperature or hot cross rolling in an attempt to investigate the effectiveness of such post-consolidation thermomechanical treatments. Its microstructure, nanoparticle dispersion and mechanical properties were analysed and correlated on each condition. As compared with the annealing treatments hot cross rolling led to a significant decrease in the grain size as well as a slight decrease in the size of some of the secondary phases present in the material. Also, the impact and tensile behaviour was improved over the whole temperature range studied.

The effect of Ar-ion irradiation on nanomechanical and structural properties of ODS RAF steels manufactured by using HIP technique

The influence of low energy ion irradiation on mechanical and structural properties of Oxide Dispersion Strengthened (ODS) Reduced Activation Ferritic (RAF) steels were investigated using Nanoindentation (NI), High Resolution Scanning Electron Microscopy (HR SEM) and Energy Dispersive X-ray Spectroscopy (EDS) techniques. ODS samples were manufactured by means of High Isostatic Pressing (HIP) technique and irradiated at room temperature with 320 keV Ar2+ ions up to fluences reaching 1 × 1016 cm−2. The results revealed fast increase of nanohardness in the function of two dependencies: (i) vanadium addition and (ii) ion fluence range. Furthermore, along with increasing vanadium content, the increase in the number of ∼100 nm sized precipitates consisting of Cr-O-V and Cr-C-V with small amounts of alloying elements (Ti, Al, W) have been observed. Simultaneously, the presence of nano-precipitates, with average size below 10 nm has been recorded. Conducted research confirms that their number decreases with increasing V content. The reported hardening effect is most probably related to: (i) quantity and size of precipitates identified as chromium-vanadium oxides and carbides with addition of alloying elements and (ii) level of damage created by ion irradiation.

Hydrogen isotopes transport in sputter-deposited tungsten coatings

For a DEMO reactor, surface coatings made of tungsten are necessary to protect the plasma-facing wall made of reduced activation ferritic steels such as F82H. In this study, hydrogen and deuterium transport parameters for sputter-deposited tungsten coatings have been evaluated by the gas-driven permeation technique in the temperature range from 200°C to 550°C. The evaluated permeability for sputter-deposited tungsten is comparable with the literature data for bulk polycrystalline tungsten. However, the apparent diffusivity and solubility are different from that of bulk polycrystalline tungsten by several orders of magnitude, which is attributed to the presence of trapping sites resulting from the characteristic microstructure of sputter-deposited tungsten coatings. The trapping effects on hydrogen migration have been discussed.