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A mechanically alloyed austenitic stainless steel (MA304LZ) was produced from pre-alloyed SUS304L powder with a small amount of Zr addition. The yield stress of MA304LZ was more than 3 times larger than that of SUS304L or 316L, while total elongation was reduced to about one third of the conventional steels. Microstructure analysis revealed an average grain size of 0.42 µm in MA304LZ and about 34/30 µm in SUS304L/316 L. In MA304LZ, two types of precipitates were observed; inhomogeneously distributed fine precipitates with an average size of 6.0 nm and homogeneously distributed coarse precipitates (d > 20 nm) with an average size of 47 nm. The strengthening mechanism of MA304LZ was discussed on the bases of Hall-Petch and Orowan equations, and the strengthening of MA304LZ was attributed mostly to refined grains. The dislocation barrier strength factor, α, is estimated to be 0.277 for the Zr-rich precipitates in MA304LZ. Keywords: Powder metallurgy, Grain size effect, Precipitation hardening, Zr-oxides

Dissimilar welding between oxide dispersion strengthened ferritic (ODS) steel and reduced activation martensitic steel would be required for constructing the advanced blanket of progressive fusion reactors. In this study, we achieved dissimilar joints by friction stir welding, and aimed to characterise and ameliorate joint inhomogeneity. Main results reveal that the joint inhomogeneity is generated from discrepant microstructural evolutions within the martensitic and ODS ferritic steels. The ODS steel achieves evolution by the dynamic recrystallisation, while the martensitic steel undergoes phase transformation that drastically hardens the stir zone. By a proper post-weld heat treatment, the joint inhomogeneity can be effectively ameliorated due to carbide reprecipitation and stress relief in the joint.

Oxide dispersion strengthened (ODS) steels can be used as the structural materials in the future fusion reactors and the fuel cladding materials in the advanced fission reactors. However, the weldability of ODS steels is a severe problem. In the present study, defect-free joints of the 15Cr-ODS ferritic steel were achieved by friction stir welding at different rotation speeds. The recrystallization, hardness and tensile properties are highly related with the rotation speed of the stir tool. The higher rotation speed results in coarser grains in the top SZ, while the grain size exhibits more complicated relation with the rotation speed in the SZ center. The joint welded at 250 rpm exhibits a maximum tensile strength of 974 MPa that reaches about 84% of that of the base metal.

ODS ferritic steels were produced by mechanical alloying (MA) in either attritor or planetary ball mills under argon atmosphere and hot extrusion. Milling in argon with controlling excess oxygen (Ex.O) and nitrogen was effective to reduce the particle size of MA powders. Coarse Al2O3 particles and fine AlN precipitates were found to be formed in an Al added steel when the Ex.O and nitrogen concentrations are increased. Lowering the Ex.O and nitrogen concentrations in ODS ferritic steels acted as a key parameter for enhancing their impact properties.

Two 14.5 wt.% Cr-ODS ferritic steels with and without simultaneous addition of 3.5 wt.% Al and 0.27 wt.% Zr were irradiated with He+ up to 0.2 dpa/3500 appm at 300 °C, 550 °C, 700 °C in order to investigate the effects of Al and Zr addition on He bubble formation behavior in 14.5Cr-ODS ferritic steel. The mean grain size of each ODS steel was about 370 nm. Al addition resulted in the formation of larger oxide particles with a lower number density in 14.5Cr-ODS ferritic steel. After irradiation to 0.2 dpa/3500 appm He, dislocation densities in both ODS ferritic steels showed no change. However, the size and number density of He bubbles were increased and decreased, respectively, by the simultaneous addition of Al and Zr. Although both ODS steels showed good resistance to He-bubble induced swelling, the 14.5Cr-ODS steel possessed a better swelling resistance than 14.5Cr-3.5 Al (Zr)-ODS steel. It was considered that the oxide particles were the main contributor for the swelling resistance of ODS steels by providing a number of defect sinks with a high sink strength. The swelling increased with increasing irradiation temperature in both ODS steels, indicating that the de-trapping He enhanced He-bubble swelling.

The growth of oxide particles in FeCrAl-oxide dispersion strengthened steel (ODSS) considering an accident condition of the light-water reactor at above 1500 K was studied by using a high-temperature annealing. Oxide particles grew from 9 nm to more than 50 nm as maximum at 1623 K for 27 h, with decreasing their number density in two orders of magnitude. Most of the oxide particles in 15Cr-7Al were identified as YAM or YAP, while the oxide particles in 15Cr-7Al-0.4Zr were identified trigonal Y4Zr3O12. Zr addition to 15Cr-7Al ODSS accelerated the growth of the oxide particles, which is quite contrary to the effect of Zr addition during sintering as suggested in the literature. The kinetics of coarsening was characterized by an equation of Ostwald ripening. The diffusion activation energies obtained in the present materials were quite larger than the conventional diffusion activation energy of Y in alpha-iron. Gibbs free energy of oxides should be considered to discuss the coarsening. (C) 2017 Elsevier B.V. All rights reserved.

Influence of the nanoscale oxide particles on mechanical properties and irradiation resistance of oxide-dispersion-strengthened (ODS) ferritic steels is of critical importance for the use of the material in fuel cladding or blanket components in advanced nuclear reactors. In the present work, impact of structures of oxide dispersoids on the irradiation hardening of ODS ferritic steels was studied. Specimens of three high-Cr ODS ferritic steels containing oxide dispersoids with different number density and average size were irradiated with high-energy Ni ions at about −50 °C. The energy of the incident Ni ions was varied from 12.73 MeV to 357.86 MeV by using an energy degrader at the terminal so that a plateau of atomic displacement damage (∼0.8 dpa) was produced from the near surface to a depth of 24 μm in the specimens. A nanoindentor (in constant stiffness mode with a diamond Berkovich indenter) and a Vickers micro-hardness tester were used to measure the hardeness of the specimens. The Nix-Gao model taking account of the indentation size effect (ISE) was used to fit the hardness data. It is observed that the soft substrate effect (SSE) can be diminished substantially in the irradiated specimens due to the thick damaged regions produced by the Ni ions. A linear correlation between the nano-hardeness and the micro-hardness was found. It is observed that a higher number density of oxide dispersoids with a smaller average diameter corresponds to an increased resistance to irradiation hardening, which can be ascribed to the increased sink strength of oxides/matrix interfaces to point defects. The rate equation approach and the conventional hardening model were used to analyze the influence of defect clusters on irradiation hardening in ODS ferritic steels. The numerical estimates show that the hardening caused by the interstitial type dislocation loops follows a similar trend with the experiment data.

The surface oxide film of a Zr−2.5Nb alloy subjected to long term corrosion at 633 K in simulated primary coolant of pressurized water reactors has been analyzed. The primary concerns were whether Nb precipitates exhibit amorphization upon oxidation, and whether they dissolve into the matrix, as suggested by previous studies. Their behavior is of particular interest, from the viewpoint of engineering, as the mechanism of improving corrosion resistance of Zr fuel cladding by Nb addition, and from the viewpoint of basic materials science, as the critical condition of solid-state amorphization. If amorphization and dissolution proceed simultaneously, it would follow that amorphization occurs at conditions where both O and Nb atoms are mobile; under such conditions diffusion-induced amorphization has never been observed. It was found that the Nb precipitates exhibited amorphization without dissolution. Some of the inconsistencies among the previous studies were found to be artifacts of materials characterization methods. The final configuration of precipitates was amorphous Nb2O5, which is distinct from the other Nb oxides in terms of its dielectric nature with a wide band gap. The matrix initially contained a large amount of Nb greater than the solubility. Although the excess Nb atoms did not precipitate by thermal aging alone, oxidation was found to enhance their precipitation at this temperature. It appears that amorphization can occur even when the motion of atoms is not frozen-in.

Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900–1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.

The age-hardening in oxide dispersion strengthened (ODS) ferritic steels with various additions of Cr (12, 15 and 18 wt.%) and Al (0, 5, 7 and 9 wt.%) were investigated. After 5000 h aging at 475 o C, the hardness increases in all these ODS steels, while the increased level depends on the Cr/Al contents. In 12Cr-ODS steels, the more the Al, the higher the increased hardness is. However, in 18Cr-ODS steels, higher Al addition suppresses the age-hardening. TEM observations of 18Cr-ODS steels reveal that 9Al suppresses the formation of Cr-enriched α' phase, while the 18Cr-5Al-ODS steel comprises a plenty of α′ phases. Adding Zr in ODS steels appears to increase the age-hardening. The susceptibility to age-hardening is remarkably lower in the ODS ferritic steels than in the non-ODS ferritic steel with the similar concentration of Cr.

The oxide particle coarsening was evaluated at temperature over 1473 K by means of transmission electron microscopy (TEM). After annealing of the 9CrODS extruded bar, the size of oxide particles increases while the number density decreases, indicating that the oxide particles coarsen through Ostwald ripening. The growth rate of the oxide particles follows the fifth-power law, which is in the region of dislocation ‘pipe’ diffusion. The activation energy for pipe diffusion, however, was remarkably high, derived as 891 KJ/mole. The stability of oxide particles and the difference of the diffusion velocity in between bcc- δ phase and fcc- γ phase should be considered as the contributions to the activation energy.

The high-temperature deformation process of the recrystallized 16CrODS ferritic steel was investigated at 1000 °C for the stress loading perpendicular to the elongated grain structure. The strain rate was varied in the range from 1.0 × 10 −2 to 1.0 × 10 − 5 s −1 . At the strain rate over 1.0 × 10 − 4 s −1 , deformation is dominated by the conventional dislocation creep. Decreasing strain rate from 1.0 × 10 − 4 s −1 , grain boundary sliding becomes prominent. Accommodation process for the localized stress induced by grain boundary sliding could be dislocation creep at 1.0 × 10 − 4 s −1 , and by diffusional creep at 1.0 × 10 − 5 s −1 or less. These were verified through the observation of void formation and localized strain accumulation by KAM map.

The effect of cold rolling on the recrystallization behavior of Al-free and Al-added 15Cr-oxide dispersion strengthened (ODS) ferritic steels was investigated. The recrystallization of both steels are enhanced by cold rolling. The Al-free ODS steel with finer oxide particles is harder to recrystallize than Al-added ODS steel with coarser oxide particles. The effect of Al addition on the recrystallization behavior is evident. It is estimated that the recrystallization temperature of Al-free and Al-added 15Cr-ODS ferritic steel is 900 &deg, C and 1250 &deg, C with the annealing period of 1 h. In Al-free ODS steel, a small hardening was observed in the temperature range between 850 &deg, C and 1200 &deg, C, while no such phenomenon was observed in Al-added ODS steel, which is indicative of retardation of recovery by finely dispersed oxide particles. Oxide particle growth is mostly dependent on annealing temperature, while recrystallization and grain growth are controlled by not only the temperature, but the cold rolling ratio, which alters the multiple factors such as dislocation density, initial grain shape and oxide particle dispersion morphology. The cold rolling direction also influences the grain morphology and grain orientation in Al-added ODS steel, and the second rolling in a perpendicular direction to the first cold rolling direction induces the rotation of the grains from &lt, 110&gt, to &lt, 112&gt, The recrystallization temperature is not significantly changed by the cold rolling direction. Recrystallization after cold rolling appears to increase the {111} grain orientation on the cold rolled specimen surface.

In the present study, friction stir welding was employed to join F82H steels and develop a potential joining technique for a reduced activation ferritic/martensitic steel. The microstructures and mechanical properties on the joint region were investigated to evaluate the applicability of friction stir welding. F82H steel sheets were successfully butt-joined with various welding parameters. In welding conditions, 100 rpm and 100 mm/min, the stirred zone represented a comparable hardness distribution with a base metal. Stirred zone induced by 100 rpm reserved uniformly distributed precipitates and very fine ferritic grains, whereas the base metal showed a typical tempered martensite with precipitates on the prior austenite grain boundary and lath boundary. Although the tensile strength was decreased at 550 °C, the stirred zone treated at 100 rpm showed comparable tensile behavior with base metal up to 500 °C. Therefore, friction stir welding is considered a potential welding method to preserve the precipitates of F82H steel.

The effects of recrystallization on ion-irradiation hardening and microstructural changes were investigated for a 15Cr-ODS ferritic steel. Dual ion-irradiation experiments were performed at 470 °C using 6.4 MeV Fe3+ ions simultaneously with energy-degraded 1 MeV He+ ions. The displacement of damage at 600 nm depth from the specimen surface was 30 dpa. Nano-indentation test with Berkovich type indentation tip was measured by constant stiffness measurement (CSM) technique. Results from nano-indentation tests indicate irradiation hardening in ODS steels even at 470 °C, while it wasn't observed in reduced activation ferritic steel. Recrystallized ODS steel shows a larger irradiation hardening, which is considered to be due to the reduction of grain boundaries and interfaces of matrix/oxide particles. In 20% cold rolled ODS steel after recrystallization, both the hardening and bubble number density were lower than those of recrystallized ODS steel, suggesting that dislocations generated by cold rolling suppress bubble formation. Based on the estimation of irradiation hardening from TEM observation results, it is considered that the bubbles are not the main factor controlling ion-irradiation hardening., Proceedings of the 19th International Conference on Ion Beam Modification of Materials (IBMM 2014)

Thermal ageing behaviors of 15 wt.% Cr-oxide dispersion strengthened (ODS) steels with the contents of Al of 5, 7 and 9 wt.%, which were isothermally aged at 475 °C for 9000 h, have been investigated by Vickers hardness measurement, tensile test and atom probe tomography (APT) to correlate the age-hardening with the evolution of nanometer-scale structures for understanding the age-hardening mechanisms. The age-hardening of 15Cr ODS steel is attributed to α−α′ phase separation. The age-hardening of 15Cr–5Al ODS steel is considerably higher, relative to 15Cr, 15Cr–7Al and 15Cr–9Al ODS steels, which is due to the strengthening by both Cr-enriched α′ phases and (Ti, Al)-enriched β′ phases formed in 15Cr–5Al ODS steel. In 15Cr–7Al and 15Cr–9Al ODS steels, however, the age-hardening is only attributed to the β′ phase strengthening because α−α′ phase separation is completely suppressed in the ODS steels when the content of Al is ≥ 7 wt.%. In all the Al-added ODS steels, fairly large amounts of β′ phases with crystallography closely related to Heusler-type Fe2AlTi precipitated during the thermal ageing. The strengthening by the β′ phases becomes more significant with the increasing of Al content from 5 wt.% to 9 wt.%. Core/shell structured oxide particles were observed only when α−α′ phase separation occurred, indicating that the formation of such core/shell structure might be closely associated with the process of α−α′ phase separation. The age-hardening mechanisms of the ODS steels with different Al-contents were proposed based on the dislocation barrier models where the amounts of age-hardening were correlated with the nanostructures characterized by APT. The theoretically predicted values of the age-hardening of the FeCrAl-ODS steels agreed well with the values experimentally determined.

The early-stage thermal age-hardening and the resultant nanoscale structure of 12Cr, 12Cr–7Al and 18Cr–9Al ODS steels have been studied by Vickers hardness measurements and atom probe tomography (APT), respectively, after isothermally ageing at 475 °C for 300 h. The thermal age-hardening of 12Cr, 12Cr–7Al and 18Cr–9Al ODS steels was measured to be 4 HV, 45 HV and 34 HV, respectively. The results of APT analyses revealed that no α–α′ phase separation occurred in all the three ODS steels, while a significant precipitation of (Ti, Al)-enriched β′ phase with crystallography closely related to Heusler-type Fe2AlTi took place in both 12Cr–7Al and 18Cr–9Al ODS steels. The chemical compositions of the β′ phase precipitates in 12Cr–7Al and 18Cr–9Al ODS steels were measured to be Fe2.3AlTi0.6Cr0.3 and Fe2.3AlTi0.5Cr0.4, respectively. In 18Cr–9Al ODS steel, no α–α′ phase separation was detected, indicating 9 wt.% Al addition into 18Cr ODS steel could intensely suppress the phase decomposition. The occurrence of β′ phase in 12Cr–7Al and 18Cr–9Al ODS steels indicates that β′ phase should be more thermodynamically stable because of its lower formation energy. The early precipitation of β′ phase may result from the higher diffusion coefficients of Al and Ti in the bcc steel matrix, relative to Cr. It was concluded that the significant early-stage thermal age-hardening in 12Cr–7Al and 18Cr–9Al ODS steels should be due to the β′ phase precipitation. The measured thermal age-hardenings agree well with the values estimated based on dislocation barrier model of Orowan mechanism.

To study the long-term thermal ageing behavior of 15Cr–2W–0.2Ti–0.25Y2O3 (15Cr) and 15Cr–4.5Al–0.5Ti–0.5Y2O3 (15Cr–5Al) ODS ferritic steels, which have been isothermally aged at 475 °C for 9000 h, the age-hardening behavior have been studied by tensile test, and the resultant nanoscale structure and chemistry have been characterized by atom probe tomography (APT), transmission electron microscopy (TEM) and energy-filtered transmission electron microscopy (EFTEM). Thermal age-hardening is much more significant in 15Cr–5Al ODS steel, relative to 15Cr ODS steel. The α′ phases are finer and less developed in 15Cr ODS steel, relative to 15Cr–5Al ODS steel. The (Ti, Al)-enriched β′ phases precipitated in 15Cr–5Al ODS steel during the ageing. Ti partitions predominantly into α′ phases in 15Cr ODS steel whereas Ti partitioning into α phases is much stronger, relative to α′ phases, in 15Cr–5Al ODS steel. The number density of core/shell structured oxide particles is much higher for 15Cr ODS steel, relative to 15Cr–5Al ODS steel. It is considered that the formation of core/shell structured oxide particles should retard α–α′ phase separation. The age-hardening of 15Cr ODS steel was well predicted by the dislocation shearing mechanism of α′ phases, while the more significant thermal age-hardening in 15Cr–5Al ODS steel is attributed to the strengthening of not only α and α′ phases but also the additional β’ phases. 15Cr ODS steel was found to be more thermally stable than 15Cr–5Al ODS steel under the present ageing conditions.

The effect of aging on high-Cr ferritic steels was investigated with focusing on the role of oxide particles in α/α′ phase decomposition behavior. 12Cr-oxide dispersion strengthened (ODS) steel, 15Cr-ODS steel and commercial SUS430 steel were isothermally aged at 475 °C for up to 10,000 h. Thermal aging caused a larger hardening in SUS430 than 15Cr-ODS, while 12Cr-ODS showed almost no hardening. A characteristic of the ODS steels is that the hardening was not accompanied by the significant loss of ductility that was observed in SUS430 steel. After aging for 2000 h, SUS430 steel shows a larger ductile–brittle transition temperature (DBTT) shift than 15Cr-ODS steel, which suggests that the age-hardening susceptibility is lower in 15Cr-ODS steel than in conventional SUS430 steel. Thermal aging leaded to a large number of Cr-rich α′ precipitates, which were confirmed by transmission electron microscopy (TEM). Correlation of age-hardening and phase decomposition was interpreted by Orowan type strengthening model. Results indicate that oxide particles cannot only suppress ductility loss, but also may influence α/α′ phase decomposition kinetics.

Oxide particle dispersion strengthened (ODS) Ni-base alloys are irradiated by using simulation technique (Fe/He dual-ion irradiation) to investigate the reliability to Gen. IV high-temperature reactors. The fine oxide particles with less than 10 nm in average size and approximately 8.0 × 1022 m−3 in number density remained after 101 dpa irradiation. The tiny helium bubbles were inside grains, not at grain-boundaries; it is advantageous effect of oxide particles which trap the helium atoms at the particle-matrix interface. Ni-base ODS alloys demonstrated their great ability to overcome He embrittlement.

The joint of oxide dispersion strengthened (ODS) ferritic steel and martensitic steel was produced by friction stir welding. The thermo-mechanical process during welding causes recrystallization in the ODS steel and phase transformation in the martensitic steel. In stir zones of both steels, low-angle boundaries significantly convert to high-angle boundaries. The recrystallization in ODS steel is characterized by the increase in the Σ3 twin boundary with an almost constant fraction of Σ5–Σ29 boundaries, while the martensitic steel shows an opposite trend.

Effects of tensile stress on Cu clustering were explained using atom probe tomography (APT) results of Fe–0.6 wt.%Cu alloy specimens irradiated with 6.4 MeV Fe ions while applying a tensile stress of 60 MPa at room temperature (less than 50 °C) and 290 °C. The hardening under the tensile-stressed irradiation was smaller than that under the stress-free irradiation at both room temperature and 290 °C. APT results showed that well-defined Cu clusters were formed in all specimens even under the room temperature irradiation. The Cu clusters under the tensile-stressed condition were smaller and had higher densities than those under the stress-free condition. The lower Cu content in clusters and more diffuse Cu clustering were obtained for the specimens irradiated under the tensile-stressed condition. The hardening efficiency of Cu clusters was correlated with the Cu content in clusters and the coherency of interface between a cluster and the matrix. Application of tensile stress would control hardening by changing the nature of Cu clusters.

The effects of thermal aging on the microstructure and mechanical properties of Fe–15Cr ferritic model alloys were investigated by TEM examinations, micro-hardness measurements and tensile tests. The materials used in this work were Fe–15Cr, Fe–15Cr–C and Fe–15Cr–X alloys, where X refers to Si, Mn and Ni to simulate a pressure vessel steel. Specimens were isothermally aged at 475 °C up to 5000 h. Thermal aging causes a significant increase in the hardness and strength. An almost twice larger hardening is required for embrittlement of Fe–15Cr–X relative to Fe–15Cr. The age-hardening is mainly due to the formation of Cr-rich α′ precipitates, while the addition of minor elements has a small effect on the saturation level of age-hardening. The correlation of phase decomposition process and age-hardening in Fe–15Cr alloy was interpreted by dispersion strengthened models.

Two kinds of ODS high-Cr ferritic steels (commercial MA956 and an Al-free 16Cr-0.1Ti ODS ferritic steel) and one conventional ferritic/martensitic steel (T122) were irradiated at about 440 degrees C with high-energy Ne-20-ions in HIRFL. Successively increasing doses from 350 to 900 appm of Ne concentration, corresponding to atomic displacement levels from 0.7 to 1.8 dpa, were approached. A nearly uniform distribution of Ne concentration and atomic displacement damage was produced through the thickness of 60 mu m of the specimens by using an energy degrader. Mechanical properties of the specimens were tested with the small-ball punch technique. The test at room temperature shows a less significant ductility loss in the ODS ferritic steel MA956 than in the T122 irradiated to the same dose of 350 appm Ne/0.7 dpa. The test at 500 degrees C shows that the Al-free 16Cr-0.1Ti ODS ferritic steel does not exhibit observable loss of ductility even to the highest dose level (900 appm Ne/1.8 dpa). An investigation with transmission electron microscopy (TEM) shows that voids with a diameter up to 70 nm were formed at grain boundaries in the conventional ferritic/martensitic steel T122 while only smaller bubbles were formed at the oxides/substrate interfaces in the ODS ferritic steel MA956. Mechanisms underlying the difference of irradiation response of the steels are discussed. (C) 2014 Elsevier B.V. All rights reserved.

The weldability of oxide dispersion strengthened (ODS) ferritic steels is a critical obstructive in the development and use of these steels. Friction stir welding has been considered to be a promising way to solve this problem. The main purpose of this work was to reveal the effects of mechanical force on grain structures of friction stir welded ODS ferritic steel. The grain appearances and the misorientation angles of grain boundaries in different welded zones were investigated by the electron backscatter diffraction (EBSD). Results showed that the mechanical force imposed by the stir tool can activate and promote the recrystallization characterized by the transformation of boundaries from LABs to HABs, and contribute to the grain refinement. The type of recrystallization in the stir zone can be classified as the continuous dynamic recrystallization (CDRX).

The stress corrosion cracking (SCC) susceptibility in supercritical pressurized water (SCPW) was investigated for the candidate structural steels of advanced fusion and fission nuclear system, which are SUS316L austenitic steel, F82H ferritic–martensitic steel and SOC-16 oxide dispersion strengthened (ODS) ferritic steel. In order to evaluate the susceptibility to SCC with those materials, slow strain rate test (SSRT) was carried out at 773 K, under a pressure of 25 MPa of SCPW with deaerated condition. High temperature tensile test in vacuum at 773 K was also performed to compare the deformation and fracture behavior between corrosive environment of SCPW and non-corrosive environment. Although SUS316L showed a change in the fracture mode in the deaerated SCPW from an entire ductile fracture at higher strain rate to a mixed mode of ductile and brittle fracture at lower one, the fracture mode of ODS steel and F82H was not changed in the tested strain rate range. Both the IGSCC and TGSCC were observed in SUS316L. And F82H steel suffered from much severer oxidation than SUS316L and SOC-16.

The effect of grain orientation, heat-treatment temperature and test temperature on the mechanical properties of tungsten (W), which vary depending on plastic working and fabrication process, was investigated by mechanical testing of tensile or bending. Heavily worked W samples (1.5–2.0 mm in the final thickness) exhibit degradation of fracture strength due to recrystallization embrittlement after heat-treatment at 1240 °C (temperature of diffusion bonding between W and a candidate material of the Fe base support structure). On the other hand, W samples with lower thickness reduction rates do not suffer degradation of fracture strength after heating up to around 1300 °C, and show somewhat higher fracture strength by heat-treatment below 1300 °C than the samples in the as-received state. The observed behavior is a reflection of recovery of dislocations introduced by plastic working. High temperature tensile testing of ITER grade W with an anisotropic grain structure and S-TUN with an equiaxed grain structure revealed that both W grades exhibit plastic elongation at temperatures higher than 200 °C with essentially the same temperature dependence of yield strength, which is relatively insensitive to grain orientation in the structure at 200–1300 °C.

F82H-IEA and its EB-weld joint were irradiated at 573 and 773 K up to 9.6 dpa and the irradiation effect on its mechanical properties and microstructure were investigated. A hardness profile across the weld joint before irradiation showed the hardness in transformed region (TR) was high and especially that in the edge of TR was the highest (high hardness region: HHR) compared to base metal (BM). These hardness distribution was correspond to grain size distribution. After irradiation, hardening in HHR was small compared to other region in the sample. In tensile test, the amount of hardening in yield strength and ultimate tensile strength of F82H EB-weld joint was almost similar to that of F82H-IEA but the fracture position of EB-weld joint was at the boundary of TR and BM. Therefore, the TR/BM boundary is the structural weak point in F82H EB-weld joint after irradiation. As the plastic instability was observed, the dislocation channeling deformation can be expected though the dislocation channel was not observed in this study.

The next generation of light water reactors, resource renewable BWR (RBWR), which can be burned trans uranium (TRU) is currently under development at Hitachi. The RBWR requires a high flux of fast neutron for efficient burning of the TRU, which is four times as large as that of the ABWR. Therefore, structural materials require both a high resistance to corrosion and to irradiation. In this study, oxide dispersion strengthened austenitic stainless steels (ODS-ASUS) with high corrosion resistance have been developed. The objective of this research is to evaluate irradiation resistance and SCC susceptibility in a simulated reactor water environment for the ODS-ASUS. The materials were irradiated with 6.4 MeV Fe3+ at 673 K up to 8.0 dpa using the DuET facility at Kyoto University. The creviced bent beam (CBB) test is conducted to assess the SCC susceptibility in the hot water, 288 °C, 8 MPa with a dissolved oxygen of 8 ppm.

For the purpose of arranging reference alloys available for various characterization efforts by Japanese fusion research groups, fabrication of reference 9Cr and 12Cr-ODS steels have been carried out with similar manufacturing processes followed by various characterizations. The fabrication proceeded with powder mixing, MA, encapsulation into mild steel cases, hot extrusion and hot forging, followed by final heat treatments. Each alloy was extruded into three bars. The characterization included chemical composition analysis, SEM and TEM microstructural observations, hardness tests, tensile tests at RT and 973 K, and relatively short-term thermal creep tests at 973 K. Room temperature hardness for 9Cr-ODS was larger than 12Cr-ODS, the former showing large increase when annealing temperature exceeded 1200 K and the latter showing no significant change with annealing temperature. Tensile strength of 9Cr-ODS was significantly larger than that of 12Cr-ODS at RT but comparable at 973 K. 9Cr-ODS showed longer and shorter creep rupture time than 12Cr-ODS at high and low stress levels, respectively. The mechanism of the difference in creep properties of the two alloys was discussed.

Bonding of oxide dispersion strengthened (ODS) steels to non-ODS reduced activation ferritic/martensitic (RAFM) steels is essential to their application to blanket systems. In the present study, a diffusion-bonded joint of the candidate 9Cr-ODS steel and JLF-1 RAFM steel was fabricated using hot isostatic pressing (HIP). The effect of post-bond heat treatments (PBHTs) was studied by hardness measurement and microstructural analysis. The results indicated that, after normalization and tempering (N&T), the hardness and microstructures of 9Cr-ODS and JLF-1 base metals recovered to levels similar to those before HIP. However, a soft region was observed across the bonding interfaces for all specimens containing the as-HIPed condition and those after PBHTs. This was due to coarser micro-carbides (M3C in as-HIPed condition and M23C6 in N&T conditions) near the interfaces than in the base metals for both 9Cr-ODS and JLF-1. Energy Dispersive X-ray Spectroscopy (EDS) analysis confirmed that carbon, tungsten, and chromium in the matrix near the interfaces are transferred to the micro-carbides, making them coarser there. Ti diffused from the 9Cr-ODS side to the JLF-1 side forming Ti-rich carbides after tempering, especially at high temperature to 1073 K.

Cavity evolutions in a normalized and tempered martensitic steel (TMS) and two nanostructured ferritic alloys (NFA) under Fe3+ and He+ dual ion beam irradiations (DII) at 500 °C and 650 °C were characterized. The irradiation conditions encompass a wide range of displacement per atom damage (dpa), He and He/dpa. The 500 °C DII produced damage and He levels of ≈10–47 dpa and ≈400–2000 appm, respectively. Transmission electron microscopy (TEM) showed that DII of a 8Cr TMS, at 500 °C to up to 60 dpa and 2100 appm He, produced a moderate density of non-uniformly distributed cavities with bimodal sizes ranging from ≈1 nm He bubbles to ≈20 nm faceted voids, and swelling ≈0.44%. In contrast, the same irradiation conditions produced only small ≈1.3 nm diameter bubbles and swelling of ≈0.05% in the NFA MA957. Similar bubble distributions were observed in MA957 and a developmental NFA DII at 650 °C up to ≈80 dpa and ≈3900 appm He. These results demonstrate the outstanding He management capability of the oxide nano-features in the NFA. The various data trends are shown as a function of dpa, He, He/dpa and He*dpa.

The friction stir welding (FSW) method was used to form joints in an oxide dispersion strengthened (ODS) steel with tool rotation speeds ranging from 250 rpm to 400 rpm to investigate the effect of rotation speed of the tool on the mechanical properties and microstructure of the joints. The hardness of the stirred zone (SZ) of the ODS steel joints was reduced after welding at all rotation speeds because of recrystallization induced by the heat generated during the FSW process. The recrystallization was accompanied by a change in the grain morphology from elongated grains to isotropic grains. The sizes of recrystallized grains increased with increasing rotation speed of the tool. Oxide particles, however, were slightly coarsened by the FSW process, and there was no systematic relationship between the particle size and the rotation speed. The hardness of the SZ of ODS steel joints followed the Hall–Petch relationship. Tensile tests of the FSW ODS steel joints revealed that the change in the ultimate tensile strength is similar to the change in the hardness, suggesting that the joint strength was reduced by grain growth but not by degradation of the fracture strength.

Stress corrosion cracking (SCC) susceptibility was investigated by means of steady strain rate tests for a 15Cr–4Al–2W oxide dispersion strengthened (ODS) ferritic steel in supercritical pressurized water (SCPW) dissolved with different hydrogen (DH) and dissolved oxygen (DO) contents. All the specimens exhibit ductile fracture mode, regardless of the strain rate. The effect of DH and DO on the fracture behavior is negligible. Small cracks were observed at necking region but most of the cracks were identified as “corrosion layer cracking” by cross-sectional observation. The ODS ferritic steel shows no susceptibility to SCC in SCPW at this experimental conditions.

The nanoparticles in an Al-alloyed high-Cr oxide dispersion strengthened (ODS) ferritic steel with Hf addition, i.e., SOC-16 (Fe-15Cr-2W-0.1Ti-4Al-0.62Hf-0.35Y2O3), have been examined by transmission electron microscopy (TEM) and high resolution transmission electron microscopy (HRTEM). Relative to an Al-alloyed high-Cr ODS ferritic steel without Hf addition, i.e., SOC-9 (Fe-15.5Cr-2W-0.1Ti-4Al-0.35Y2O3), the dispersion morphology and coherency of the oxide nanoparticles in SOC-16 were significantly improved. Almost all the small nanoparticles (diameter 10 nm) were also mainly identified as cubic Y2Hf2O7 oxides with the anion-deficient fluorite structure. The results presented here are compared with those of SOC-9 with a brief discussion of the underlying mechanisms of the unusual thermal and irradiation stabilities of the oxides as well as the superior strength, excellent irradiation tolerance and extraordinary corrosion resistance of SOC-16.

Joint strengthening of oxide dispersion strengthened (ODS) martensitic steel has been attained by a transient liquid phase (TLP) bonding utilizing a newly developed ODS insert foil. The ODS insert foil (Fe–9Cr–2W–0.2Ti–0.35Y 2 O 3 –0.5C–3B–2Si) was fabricated using mechanical alloying and a spark plasma sintering method. Compared to conventional TLP bonding with a non-ODS insert foil (Fe–0.5C–3B–2Si), the microstructure of the melted zone consists of finer grains in the joint with the newly developed ODS insert material, and the grain size is almost one third of that in the conventional insert material. This increases the hardness by ΔHv = 100 in the region of the joints. Oxide particles that are coherent to the ferritic matrix in the ODS insert foil could be responsible for the grain refinement, which is explained in terms of enhanced nucleation of consolidation matrix at the oxide particles.

A δ-Zr-hydride was irradiated with 6.4 MeV Fe3+ ions to clarify the relationship between hardening and microstructural changes of bulk Zr-hydrides under neutron irradiation. Irradiation hardening was measured by nanoindentation tests. Transmission electron microscope cross-sectional observations showed that the deformation mechanism of the δ-Zr-hydride was both slip and twinning. Dislocation loops were observed in the irradiated hydride matrix. These irradiation-induced defects make slip deformation difficult and consequently promote the twin deformation of δ-Zr-hydride. This work is a continuation of the previous our work (J. Nucl. Mater. 419 (2011) 366–370) focused upon e-Zr-hydride and we discuss a comparison between the two Zr-hydrides.

Nano-bubble dispersion strengthened (N-BDS) Fe was made from Fe and polymethylmethacrylate (PMMA) powder and irradiated by 6.4 MeV Fe 3+ ions to investigate the cavity strengthening and the bubble to void evolution. The bubbles accelerated the irradiation-induced cavity growth. The hardness of the N-BDS Fe was 500 MPa higher than that of unalloyed Fe and the hardness increased by irradiation, while that of unalloyed Fe did not increase. Cavity is probably the origin of the irradiation hardening of N-BDS Fe.

Fracture toughness data was obtained for reduced-activation ferritic (RAF) steels with different sizes of specimens (1 compact tension(CT), 1/2 CT and 1/4 CT) using the master curve (MC) method in the transition temperature region. Considering the size adjustment by ASTM E1921, effects of specimen size on the fracture toughness are not observed and the reference temperature (T0) is around 164 K which is similar to those (154 K) of other previous studies. However, the data are not well represented by a MC, showing a rather large number of data below the lower boundary curve. Our proposed new MC was derived within the framework of the ASTM E1921 standard to apply the MC method to F82H steel. This new MC analysis can be applied to RAF steels to estimate T0 with a better description of the data scatter in the transition temperature region of fracture toughness than that of the conventional MC analysis.

Several types of reduced activation ferritic/martensitic (RAFM) steel have been developed over the past 30 years in China, Europe, India, Japan, Russia and the USA for application in ITER test blanket modules (TBMs) and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical properties of these steels before and after irradiation and in contact with different cooling media. This paper presents recent RAFM steel results obtained in ITER partner countries in relation to different TBM and DEMO options. (C) 2012 Elsevier B. V. All rights reserved.

Ion irradiations with 6.4 MeV Fe 3+ were performed on pure-Fe, Fe–1 at.% Cr, Fe–1 at.% Mn, and Fe–1 at.% Ni to a nominal damage of 1 dpa, at a damage rate of 1 × 10 −4 dpa/s, at irradiation temperatures of 473, 563, and 673 K. After irradiations at 473 and 563 K, Fe–1Mn and Fe–1Ni showed significant irradiation hardening, which was due to irradiation induced dislocation loops in high density. In pure-Fe, the dislocation loops were localized in the vicinity of dislocations, while those in Fe alloys were distributed rather homogeneously. This can be interpreted in terms of the interaction between alloying element and dislocation strain field. Irradiation at 673 K resulted in the formation of voids in pure-Fe, Fe–1Cr, and Fe–1Mn. We found that chromium suppressed void swelling.

The irradiation hardening behavior of reduced-activation ferritic steels after single Fe-ion beam irradiation and dual-ion (Fe ion and He ion) beam irradiation experiments was investigated with nanoindentation tests. The ion-irradiation experiments were conducted at 563 K with 6.4 MeV Fe3+ ions up to 3 dpa at a 600 nm depth from the irradiated surface. Furthermore, these experiments were conducted with and without simultaneous energy-degraded 1 MeV He+ ions up to 300 appm. The materials used were F82H, F82H + 1Ni, and F82H + 2Ni to investigate the effect of Ni addition on the irradiation hardening behavior. The measured nanoindentation hardness was converted to the bulk-equivalent hardness based on a combination of the Nix–Gao model to explain the indentation size effect and the composite hardness model to explain the softer substrate effect of the nonirradiated region beyond the irradiated depth range. It is clearly shown that the Ni addition enhances the irradiation hardening of F82H. The bulk-equivalent hardness is compared with the experimentally obtained Vickers hardness of F82H steels after neutron irradiation. The effect of simultaneously implanted helium on the irradiation hardening is negligible in the investigated irradiation conditions.