Search

The nickel based alloy Inconel 617 is a leading candidate for use in Generation IV fission reactors due to its superior high-temperature properties. The hardening and microstructural evolution in this alloy under irradiation with 4 MeV Xe20+ ions up to 10 dpa were characterized herein to reveal the irradiation resistance of Inconel 617. The relationship between irradiation induced microstructural evolution and the changes to the nanohardness calculated by the volume fraction model (VFM) was established. The results revealed dislocation loops are the major hardening source. A comparative analysis with GH3535 alloy revealed that the much higher number density of dislocation loops in Inconel 617 led to inferior hardening resistance than that of GH3535 alloy after irradiation at room temperature at doses below 10 dpa. However, the hardening degree of both alloys is nearly equivalent when the irradiation dose increases over 10 dpa.

Investigation of microstructures and hardening property of initial irradiated, post-irradiation annealed, and re-irradiated Chinese-type low-Cu (0.01 wt%) reactor pressure vessel (RPV) steel were conducted by slow positron beam (SPB), TEM, and nanoindentation. Results of the SPB measurements indicated that a large number density of open volume defects such as vacancies, vacancy clusters, vacancy-solute/H complexes, and dislocation loops were introduced in both initial-irradiated and re-irradiated specimens. The TEM results indicated that interstitial-type dislocation loops with a number density of ∼1022 m−3 were formed and the mean size (∼3 nm) of the dislocation loops was almost unchanged in both initial-irradiated and re-irradiated specimens. The open volume defects such as vacancy-type defects and dislocation loops were almost annealed out and some stable defects still existed in the post-irradiation annealed specimen. The nanoindentation results identified that the obvious hardening phenomena were found in the initial irradiated, post-irradiation annealed and re-irradiated specimens. The quantitative analysis suggested that irradiation hardening of Chinese-type low-Cu RPV steel were attributed to matrix defects including large-size vacancy clusters, vacancy-solute/H complexes, and dislocation loops produced during initial irradiation and re-irradiation. On the other hand, the existed stable defects such as solute-atom (Mn, Ni, and Si) complexes/precipitates, which cannot be distinguished by SPB or TEM, might be responsible for the rest fraction of irradiation hardening of highly initial irradiated and re-irradiated Chinese-type low-Cu RPV steel.

The ferritic/martensitic steel HT9 was irradiated in the BOR-60 reactor at 650, 690 and 730 K (377, 417 and 457 °C) to doses between ∼14.6–18.6 displacements per atom (dpa). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the influence of irradiation temperature on microstructure evolution. Mn/Ni/Si-rich (G-phase) and Cr-rich (αʹ) precipitates were observed within martensitic laths and at various defect sinks at 650 and 690 K (377 and 417 °C). For both G-phase and αʹ precipitates, the number density decreased while the size increased with increasing temperature. At 730 K (457 °C), within martensitic laths, a very low density of large G-phase precipitates nucleating presumably on dislocation lines was observed. No αʹ precipitates were observed at this temperature. Both a and a/2 type dislocation loops were observed, with the a type being the predominant type at 650 and 690 K (377 and 417 °C). On the contrary, very few dislocation loops were observed at 730 K (457 °C), and the microstructure was dominated by a/2 type dislocation lines (i.e., dislocation network) at this temperature. Small cavities (diameter 2 nm) were observed only at 690 K (417 °C), resulting in a bimodal cavity size distribution at 690 K (417 °C) and a unimodal size distribution at 650 and 730 K (377 and 457 °C). The highest swelling (%) was observed at 690 K (417 °C), indicating that the peak of swelling happens between 650 and 730 K (377 and 457 °C).

China Low Activation Martensitic (CLAM) steel as one of 9%Cr ferritic-martensitic steels was designed as the structural materials for fusion reactors and lead-based Reactor in China. The long-term thermal stability of precipitated phases in CLAM steel at 550 °C was systematically studied in this paper, including coarsening kinetics, number density and shape stability. Based on large data statistics, the coarsening kinetics of M23C6 carbide, MX carbonitride and Laves phase were analyzed to develop a coarsening model of each precipitated phase, respectively. The results showed that the coarsening rate of Laves phase was about 100 times than that of M23C6 carbide after aging for 20,000 h at 550 °C. The number density of precipitated phases increased firstly and then decreased with increasing of aging time due to a dynamic equilibrium process of the precipitation and dissolution of precipitated phase. The results will help to deep understand degradation mechanisms after thermal aging.

In-situ heavy ion (1 MeV Kr2+) irradiation was carried out to investigate the evolution of irradiation induced lattice defects at room temperature and the stability of β-Zr under irradiation at temperatures ranging from 150 to 350 °C. It was found that the irradiation induced defects started to appear at a dose between 0.5 dpa and 0.75 dpa, which is relatively high compared to the case for α-Zr. The number density of dislocation loops increased exponentially with fluence, with an exponential index ∼3.52. The statistical results show that the electron beam with which in-situ observation was carried out had an influence on the size of the dislocation loops. The g·b analysis shows that dislocation loops with Burgers vector ½ 1 ¯ 11> and are both present in the sample after irradiation to 1.5 dpa. ω phase particles precipitate out during irradiation at 250 and 350 °C, as a result of the combination of the thermal treatment and irradiation enhanced concentration of point defects.

A series of Fe-3 to 18 at.% Cr binary ferritic alloys were neutron irradiated side by side, in the Advanced Test Reactor (ATR) at a temperature of ∼320 °C to a dose of ∼1.8 dpa. Three types of features that form under irradiation are relevant: (i) solute clusters; (ii) α′ precipitates; and, (iii) dislocation loops. The size and number density of the precipitates and loops were measured by transmission electron microscopy (TEM) and compared to previous atom probe tomography (APT) observations. The loop density systematically decreases with increasing Cr, while the number of α′ precipitates increase at 9Cr and above. A major objective of this work is to estimate the dispersed barrier obstacle strength factors (αj) for loops, clusters and α′ precipitates, based on the combined microstructural observations and corresponding irradiation hardening measurements. Standard dispersed barrier-hardening models and computationally derived superposition rules were least square fit to determine the αj. The optimized hardening predictions are in very good agreement with experiment if mixed linear sum and root square sum superposition rules are used. Five increments of 168 h isochronal anneals of the 6Cr alloy between 300 and 400 °C coarsened the loops, while 300 h anneals of the 18Cr at 500 and 600 °C coarsened and dissolved the α′ precipitates, respectively, consistent with the Fe-Cr phase diagram.

Two kinds of 9Cr 2W steels with and without the addition of 0.1 wt% Si were irradiated with neutrons up to a fluence of 4.8 × 1023 n/m2 at 563 K. Precipitates, dislocation loops, radiation-enhanced nano-clusters and radiation-hardening were studied using advanced electron microscopy and micro-hardness test. Si reduced the number density of dislocation loops, and therefore the extent of radiation-hardening decreased. It also favored the formation of silicide-like nano-clusters. In this study, the role of Si in neutron irradiated 9Cr 2W steels is discussed, and then extended to nano-clusters and irradiation-hardening.

The Hastelloy N alloy were irradiated by He ions, Xe ions and Xe + He ions at 650 °C and examined using transmission electron microscopy (TEM) and nano-indentation. TEM results showed the formation of irradiation induced helium bubbles, dislocation loops and precipitates in the irradiated samples. The mean size, number density and their variations of these defects at the peak damage regions were measured through TEM micrographs. The interactions between different types of irradiation induced defects were investigated and the mechanisms behind them have been revealed in this study. In addition, the nanoindentation test showed the irradiation induced hardening in all the irradiated samples. The measured nanohardness increment (ΔH) for sample irradiated by Xe + He ions was larger than that of the combined single ion irradiation case, indicating the presence of synergistic effect in the hardening of the both ions irradiated sample.

This study explores the evolution processes of initiation, migration, growth, annihilation, and aggregation for dislocation loops and dislocation complex formation in regions with and without pre-existing dislocations. In-situ TEM experimental observations are conducted for 508-III steel, commonly used as a pressure vessel steel in nuclear reactors, during 400 keV Fe+ irradiation at 573 K. Loop initiation is always observed, even following the formation of dislocation complex. The average size and number density of dislocation loops, in regions both with and without pre-existing dislocations, are obtained as a function of the irradiation dose. At the initial stage of irradiation, pre-existing dislocations significantly influence the evolution of dislocation loops. However, with an increase in the irradiation dose, this influence gradually weakens. The in-situ TEM experimental observations of microstructure evolutions in this study are of significant value for furthering our understanding of irradiation damage to 508-III steel.

The loop evolution in Mo-5wt.%Re alloy was studied by in-situ TEM observation under 400 keV Fe+ irradiation at 700 °C and 800 °C. Loop initiation always existed during the whole irradiation process. The gradual disappearance of dislocation line to grain boundary and the change of loop habit plane from [101] to [211] were in-situ observed, and the changed loop acting as an intermediate bridge led to the merging of three loops. Three main ways causing loop disappearance including the aggregation between loops, the absorption by strong defect sinks and the influence of surrounding loops were obtained from in-situ experiment. With the increase of irradiation dose, many loops formed in coarse grains but few in nanocrystals, and the areal density of dislocation lines almost kept stable in the grain with many pre-existing dislocation lines. The variation trends and relationships of the average size and volume number density of loops with irradiation temperature, irradiation dose and sample thickness were obtained. At the same irradiation temperature, the thicker the thickness of TEM foil was, the higher the yield strength increasement would be. Meanwhile, for the same sample thickness, higher temperature would reduce the irradiation hardening effect. The in-situ observation of microstructure evolution will be helpful to deeply understand the irradiation damage behavior of Mo-Re alloy.

Pure iron was irradiated with 6.4 MeV Fe3+ ions at 300 °C to a displacement damage of 30 dpa at a nominal depth of 600 nm. Distinct microstructures were observed at different depth regions by transmission electron microscopy (TEM). In the range of 0–1.2 μm from the irradiation surface, dislocation loops tended to form ordered loop rafts with their habit plane on {100}, while in the range of 1.6–2.3 μm, both line-shaped dislocations and dislocation loops coexisted. Dislocation loops that formed the loop raft had an average size of 3.5 nm with the number density estimated to be 3.6 × 1023 m−3, considering all three variants of the Burgers vectors b = a0 . However, the dislocation loops in the range of 1.6–2.3 μm were with Burgers vectors, b = a0/2 , and the number density was one order of magnitude lower.

The microstructural evolution of the 21Cr32Ni model alloy under ion irradiation is investigated. A set of bulk materials were irradiated at the Michigan Ion Beam Laboratory using single beam (5 MeV Fe++) to 1, 10 and 20 dpa at 440 °C and dual beam (5 MeV Fe++ plus energy degraded 1.95 MeV He++ ions) to 16.6 dpa at 446 °C. The average diameter and number density of the faulted loops and cavities formed under irradiation were characterized using Transmission Electron Microscopy (TEM). The behavior of faulted loop in the model alloy was also investigated in-situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory (ANL). Results show that the average faulted loop diameter decreases, but the faulted loop number density increases with increasing dose. In-situ experiments showed that the faulted loops become unfaulted during ion irradiation by interacting with network dislocations. Although the average faulted loop diameter after 16.6 dpa dual beam irradiation at 446 °C was found to be similar to those seen in samples irradiated with single beams to 10 and 20 dpa, the faulted loop number density was significantly higher in the dual beam irradiated sample. Moreover, the dual beam irradiated model alloy exhibits a significantly higher density of smaller cavities. It is also found that the size and density of the faulted loops and voids calculated for the dual beam irradiation of 21Cr32Ni model alloy at 446 °C are in better agreement with those measured in a sample neutron irradiated at 375 °C. Further discussion is presented in this study.

In the present work, a numerical model based on the coupling of Kampmann and Wagner Numerical (KWN) framework and thermodynamic software ThermoCalc has been developed to predict grain boundary precipitate evolution in 316H stainless steel during thermal aging. The model is calibrated and validated against precipitate size distributions obtained by accelerated isothermal heat treatment and analysed using scanning electron microscopy (SEM). Elemental distribution was also investigated using electron microprobe analysis (EPMA). The predicted average particle size, particle size distribution and precipitate number density predicted by the model were found to be in good agreement with the experimental results. The model was then applied to predict the particle size distribution after several years exposure at service temperature. It is demonstrated that these predictions are consistent with measurements from a service-exposed part. The sensitivity of the precipitate size distribution to temperature is emphasised, and it is demonstrated that the model has potential as a useful tool for predicting evolution of the precipitate size distribution during service, providing reliable thermal data are available for the whole service life.

FeCrAl-ODS ferritic steels with Ce-oxide dispersion instead of Y-oxide were produced for the accident tolerant fuel cladding of the light water reactor. Excess oxygen (Ex.O) was added to improve the mechanical property. The tensile strength at Ex.O = 0 is around 200 MPa at 700 °C, mainly owing to dispersed Ce2O3 particles in less than 10 nm size. The formation of the fine Ce2O3 particles is dominated by a coherent interface with ferritic matrix. With increasing Ex.O, an increased of number density of coarser Ce-Al type oxide particles over 10 nm size is responsible for the improvement of the tensile strength. Change of the type of oxide particle, CeO2, Ce2O3, CeAlO3, Al2O3, in FeCrAl-ODS steel was thermodynamically analyzed as a parameter of Ex.O.

In order to investigate the effects of oxide particles on radiation response such as hardness change and microstructural evolution, three types of oxide dispersion strengthened (ODS) ferritic steels (named Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS), mostly strengthened by Y-Ti-O, Y-Al-O and Y-Zr-O dispersoids, respectively, were simultaneously irradiated with iron and helium ions at 550 °C up to a damage of 30 dpa and a corresponding helium (He) concentration of ∼3500 appm to a depth of 1000–1300 nm. A single iron ion beam irradiation was also performed for reference. Transmission electron microscopy revealed that after the dual ion irradiation helium bubbles of 2.8, 6.6 and 4.5 nm in mean diameter with the corresponding number densities of 1.1 × 1023, 2.7 × 1022 and 3.6 × 1022 m−3 were observed in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS, respectively, while no such bubbles were observed after single ion irradiation. About 80% of intragranular He bubbles were adjacent to oxide particles in the ODS ferritic steels. Although the high number density He bubbles were observed in the ODS steels, the void swelling in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS was still small and estimated to be 0.13%, 0.53% and 0.20%, respectively. The excellent swelling resistance is dominantly attributed to the high sink strength of oxide particles that depends on the morphology of particle dispersion rather than the crystal structure of the particles. In contrast, no dislocation loops were produced in any of the irradiated steels. Nanoindentation measurements showed that no irradiation hardening but softening was found in the ODS ferritic steels, which was probably due to irradiation induced dislocation recovery. The helium bubbles in high number density never contributed to the irradiation hardening of the ODS steels at these irradiation conditions.

Transmission electron microscopic (TEM) in situ mechanical testing is a promising method for understanding plasticity in shallow ion irradiated layers and other volume-limited materials. One of the simplest TEM in situ experiments is cube-corner indentation of a lamella, but the subsequent analysis and interpretation of the experiment is challenging, especially in engineering materials with complex microstructures. In this work, we: (a) develop MicroViBE, a motion detection and background subtraction-based post-processing approach, and (b) demonstrate the ability of MicroViBe, in combination with post-mortem TEM imaging, to carry out an unbiased qualitative interpretation of TEM indentation videos. We focus this work around a Fe-9%Cr oxide dispersion strengthened (ODS) alloy, irradiated with Fe2+ ions to 3 dpa at 500 °C. MicroViBe identifies changes in Laue contrast that are induced by the indentation; these changes accumulate throughout the mechanical loading to generate a “heatmap” of features in the original TEM video that change the most during the loading. Dislocation loops with b = ½ identified by post-mortem scanning TEM (STEM) imaging correspond to hotspots on the heatmap, whereas positions of dislocation loops with b = do not correspond to hotspots. Further, MicroViBe enables consistent, objective quantitative approximation of the b = ½ dislocation loop number density.

The description of intra-granular fission gas behaviour is a fundamental part of any model for the prediction of fission gas release and swelling in nuclear fuel. In this work we present a model describing the evolution of intra-granular fission gas bubbles in terms of bubble number density and average size, coupled to gas release to grain boundaries. The model considers the fundamental processes of single gas atom diffusion, gas bubble nucleation, re-solution and gas atom trapping at bubbles. The model is derived from a detailed cluster dynamics formulation, yet it consists of only three differential equations in its final form; hence, it can be efficiently applied in engineering fuel performance codes while retaining a physical basis. We discuss improvements relative to previous single-size models for intra-granular bubble evolution. We validate the model against experimental data, both in terms of bubble number density and average bubble radius. Lastly, we perform an uncertainty and sensitivity analysis by propagating the uncertainties in the parameters to model results.

Post-irradiation annealing was performed on a 304L SS that was irradiated to 5.9 dpa in the Barseback 1 BWR reactor. Evolution of dislocation loops, radiation-induced solute clusters and radiation-induced segregation at the grain boundary was investigated following thermal annealing at 500 °C and 550 °C up to 20 h. Dislocation loops, Ni-Si and Al-Cu clusters, and enrichment of Ni, Si and depletion of Cr at the grain boundary were observed in the as-irradiated condition. Dislocation loop size did not change significantly after annealing at 550 °C for 5 h but the loop number density decreased considerably and loops mostly disappeared after annealing at 550 °C for 20 h. The average size of Ni-Si and Al-Cu clusters increased while the number density decreased with annealing. The increase in cluster size was due to diffusion of solutes rather than cluster coarsening. Significant volume fractions of Ni-Si and Al-Cu clusters still remained after annealing at 550 °C for 20 h. Substantial recovery of Cr and Ni at the grain boundary was observed after annealing at 550 °C for 5 h but neither Cr nor Ni was fully recovered after 20 h. Annihilation of dislocation loops, driven by the thermal vacancy concentration gradient caused by the strain field and stacking fault associated with the loops appeared to be faster than annihilation of solute clusters and recovery of Ni and Si at the grain boundary, both of which are driven by the solute concentration gradients.

Single crystal cubic phase silicon carbide (3C-SiC) films on Si were implanted to 9.6 × 1016 25Mg+/cm2 at 673 K and annealed at 1073 and 1573 K for 2, 6, and 12 h in an Ar environment. The data from scanning transmission election microscopy (STEM) and electron energy loss spectroscopy (EELS) mapping suggest a possible formation of unidirectionally aligned tetrahedral precipitates of core (MgC2)-shell (Mg2Si) in the implanted sample annealed at 1573 K for 12 h. There are also small spherical voids near the surface and larger faceted voids around the region of maximum vacancy concentration. Atom probe tomography confirms 25Mg segregation dominated by small atomic clusters with local 25Mg concentrations up to 85 at.%. The resulting precipitate size and number density are found to decrease and increase, respectively, probably as a result of the thermal annealing that decomposes the 25Mg-bearing precipitates at the elevated temperatures and subsequent nucleation and growth below 1073 K during the cooling stage. The results from this study provide data needed to fully understand the property degradation of SiC in a high-flux fast neutron environment.

Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 × 1020 n/cm2 were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Mn, Ni, and Si atoms like a core-shell structure. In low-Cu bearing materials, Mn, Ni, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius ( r ) and number density ( N d ) decreased and increased with increasing nominal Si content, respectively. The shift in the reference temperature for nil-ductility transition (ΔRTNDT) showed a good correlation with the square root of volume fraction ( V f ) multiplied by ( V f × r ) . The negative relation between the nominal Si content and ΔRTNDT indicated that increasing of nominal Si content reduces the degree of embrittlement.

The microstructural evolution of ternary Fe-21Cr-32Ni (21Cr32Ni) model alloy and of alloy 800H was investigated with a series of in-situ ion irradiation experiments performed using the Intermediate Voltage Electron Microscope (IVEM)-Tandem Facility at Argonne National Laboratory (ANL). Samples were irradiated in-situ with 1 MeV Kr++ ions in the temperature range of 50K to 713K to doses up to 2 dpa. The size distribution of defect clusters, average defect cluster diameter, and defect cluster density were measured and compared. Results showed that the evolution of defects (i.e. the average defect cluster size and number density) in 21Cr32Ni model alloy and alloy 800H with dose were similar at irradiation temperatures up to 300K where they initially increased with dose up to 0.1 dpa after which no significant changes in defect size and density were observed with further irradiation. In addition, both alloys exhibited ordered defect structures along the direction at relatively low temperatures, up to 300K, which remained stable throughout post-irradiation in-situ thermal annealing up to a temperature of 773K. During irradiation at 713K, small defect clusters were observed at low doses (

Dislocation loops finely disperse in bulk W are generally visible to the transition electron microscopy (TEM) after irradiation. In the absence of strong interactions, these loops would normally diffuse very fast until being sunk at grain boundaries or at the dislocation network. In this work, we evaluate the strength of two pining effects that can explain the reason why they are nevertheless observed by TEM in bulk. On the one hand, we evaluate with density functional theory (DFT) the strength of binding between isolated loops and dissolved chemical impurities. Employing classical equations of diffusion, we estimate the resulting effective diffusion coefficient of loops. On the other hand, we consider the effect of mutual elastic interactions (MEI) between the loops, applying linear elasticity. We perform a large set of kinetic Monte Carlo (KMC) simulations, aimed at evaluating the effective diffusion coefficient, accounting for multiple interactions. Finally, we draw a map that indicates the dominant pinning effect given the experimental conditions (loop size and loop number density). Comparing with a large database of experimental TEM evidence, we conclude that pinning by dissolved impurities is the dominant effect.

We have developed a new embrittlement trend curve (ETC) for neutron-irradiated reactor pressure vessel(RPV) steels based on Japanese surveillance and material test reactor data. Our model is based on experimental data obtained by atom probe tomography (APT) and Charpy V-notch impact tests. The square root of the volume fraction (Vf) of the solute atom cluster (SC) formed by neutron irradiation, which is determined by APT, is correlated to the Charpy transition temperature shift (ΔT41J). In addition to the ΔT41J and Vf data, other APT data (the average volume of solute clusters (SC), the number density of SC, and the matrix Cu content in RPV steels) were also used to calibrate the coefficients in the new ETC model. The new ETC model has a simple functional form and, unlike the current version of JEAC4201, does not require numerical time integration. The new ETC model has similar ΔT41J prediction performance to the current JEAC4201 model. Furthermore, the APT data are also predicted well, which suggests a good physical basis for the new ETC.

One of the key embrittlement mechanisms in reactor pressure vessel (RPV) steels is the hardening produced by nanometer features. In this paper low Cu high Ni VVER-1000 RPV steels with a wide range of Ni (1.1–1.94wt.%) and Mn (0.38–1.1wt.%) contents irradiated at 290°C to (6–101) 1022 n/m22 (E>0.5MeV) at both low (surveillance specimens) and high (test reactor) fluxes were analyzed using atom probe tomography and transmission electron microscopy. Formation of high number density of NMS precipitates with the average composition close to at.%: 45Ni-33Si-22Mn regardless of the irradiation conditions was observed. Its number density increases with the increase of the total Ni and Mn concentration in steel. High flux irradiation demonstrates lower size and volume fraction of NMS precipitates compared to the low flux irradiation in case of welds. For the base metal specimens with low Ni (1.1–1.3 wt.%) and Mn (0.38–0.51) content no flux effect was observed. The contribution of NMS phases and dislocation loops into radiation hardening of RPV steels was established by the dispersed barrier strength model (Orowan equation).

NF616 is a third-generation ferritic martensitic steel, developed to have better creep resistance than the prior generation T91. Unlike relatively numerous studies of T91, there is a lack of understanding of the irradiation effect on the microstructural evolution and mechanical response of NF616. This work evaluated the microstructures and radiation hardening of NF616 irradiated up to 8.2 displacement per atom (dpa) at 292 °C – 431 °C, compared with T91 from two heats. Dislocation loops were observed in all investigated samples. NF616 exhibited comparable loop size but slightly lower loop density than those in the general T91 heat at 430 °C. Cavities were only observed in NF616 at 431 °C but absent at lower irradiation temperatures (292 °C and 359 °C). Ni-rich clusters were also observed in NF616 at 431 °C, while only weak Ni-clustering were observed at lower irradiation temperatures. Compared to the general T91 heat, NF616 demonstrated better swelling resistance (e.g., one third of swelling in the general T91 heat at 430 °C), a slightly higher number density of Ni-rich clusters, and slightly lower radiation hardening. The low-carbon T91 showed the greatest hardening with the largest swelling and loop sizes, despite its lowest irradiation temperature and intermediate dose. The calculated hardening from loops, cavities and Ni-rich clusters using the classic dispersed barrier-hardening model had reasonable agreement with the experiment-derived results, with the primary hardening contribution attributed to dislocation loops.

To evaluate the effects of irradiation on additive manufacturing materials which are expected to be applied in advanced nuclear energy systems, selective laser melting 316L stainless steel (SLM 316L SS) and traditionally manufactured 316L stainless steel (TM 316L SS), were irradiated with 500 keV He ions at 700°C, respectively. SEM, TEM and nanoindentation measurements were performed to survey the effects of helium (He) bubbles on the irradiation hardening. It was found that the degree of irradiation hardening of the two materials increased with an increase of ions fluences. The hardening range of SLM 316L SS was 6-47% while that of TM 316L SS was 32-75%. Moreover, the hardening saturation phenomenon was more pronounced in TM 316L SS than in SLM 316L SS. These results suggested that SLM 316L SS owned higher resistance to irradiation hardening than TM 316L SS. TEM observations showed that the density of He bubbles in TM 316L SS was 7.06 × 1023/m3 while that in SLM 316L SS was 3.33 × 1023/m3. Further, the average sizes of bubbles were 2.2 and 2.4 nm, respectively. The differences of the number density might be why SLM 316L SS had a slighter irradiation hardening than TM 316L SS, which was consistent with Orowan model. It was suggested that the numerous intrinsic microstructures such as entangled dislocation network walls and subgrain boundaries in SLM 316L SS could greatly reduce the density of He bubbles.

The objective of this study is to evaluate the influence of dose rate and cascade morphology on nanocluster evolution in a model Fe-9%Cr oxide dispersion strengthened steel and the commercial ferritic/martensitic (F/M) alloys HCM12A and HT9. We present a large, systematic data set spanning the three alloys, three irradiating particle types, four orders of magnitude in dose rate, and doses ranging 1–100 displacements per atom over 400–500 °C. Nanoclusters are characterized using atom probe tomography. ODS oxide nanoclusters experience partial dissolution after irradiation due to inverse Ostwald ripening, while F/M nanoclusters undergo Ostwald ripening. Damage cascade morphology is indicative of nanocluster number density evolution. Finally, the effects of dose rate on nanocluster morphology provide evidence for a temperature dilation theory, which purports that a negative temperature shift is necessary for higher dose rate irradiations to emulate nanocluster evolution in lower dose rate irradiations.

The initial phase separation and coarsening of the Cr-enriched α′ phase in Fe–Cr alloys were studied by utilizing phase field model. The phase separations of the α′ phase via non-classical nucleation and growth, the transition state from nucleation and growth to spinodal decomposition, and the spinodal decomposition were investigated at 700 K for the alloys containing 25, 28 and 35 at.% Cr. The time exponent of the average radius of α′ phase decreases from the growth and coarsening stage to the steady-state coarsening stage for the nucleation growth mechanism, the slope of number density of α′ phase increases at the steady-state coarsening stage with the increased Cr concentration. The velocity of phase separation in 25 at.% Cr alloy with a nucleation growth mechanism are slower than that in high Cr alloys with a spinodal decomposition mechanism. The results indicate that a low concentration alloy is more stable at high temperature, and the phase separation is fast at the early stage, which supply a reference for the alloy composition selection, and for the analysis of the relationship of property and morphology.

Bulk samples of the GH3535 alloy weld metal have been characterized by transmission electron microscopy, X-ray diffraction and nanoindentation to determine their microstructural evolution and mechanical property changes after 8 MeV Ni 3+ ions irradiation. The irradiation experiments were carried out at room temperature and 600 °C, and the ion fluences correspond to a calculated peak damage dose of 0.5, 2 and 12 dpa, respectively. TEM results show the formation of solute clusters or dislocation loops with a number density of approximately 5–14 × 10 22 m −3 and sizes between 3 and 10 nm at room temperature due to irradiation induced defects and their evolution. Moreover, the peak shift with increasing ion dose observed in XRD diffraction patterns reveals the lattice distortion induced by ion irradiation. As far as the case of high temperature irradiation, several solute clusters with the same size were observed, whereas the number density was smaller than that of the former case. The calculated indentation values in irradiated samples were found to be much higher in comparison to the unirradiated one, indicating the dose dependent effect of irradiation on hardness. However, in the case of the ion irradiation at 600 °C, the hardness value of samples was significantly decreased. The relationship between ion irradiation induced microstructural evolution and the changes in the mechanical properties of this weld metal is discussed in the context of ion dose and irradiation temperature.

We have assessed the formation and evolution of void and dislocation arrangements in Ta, Ta-5wt.%W and Ta-10wt.%W as a function of radiation level at a temperature of 345 ± 3 °C, by combining proton irradiation experiments, transmission electron microscopy and nano-hardness measurements. The damaged structure of tantalum at 0.1 dpa is characterized by the presence of a /2 〈111〉 interstitial dislocation loops and randomly distributed voids, whereas only dislocation loops are observed in the two alloys. Void ordering occurs in tantalum at 0.25 dpa, together with the appearance of dense dislocation tangles. A further increase in damage level leads to a continuous nucleation and growth of voids, and saturation is not attained at a damage level of 1.55 dpa. In contrast, the average size and number density of dislocation loops increases gradually with damage level in the two alloys, and voids only form at 1.55 dpa. Tungsten delays the loop evolution and therefore the formation of radiation-induced voids.

A nanostructured ferritic steel with nominal composition of Fe-14Cr-1Ti-0.3Mo-0.5La 2 O 3 (wt.%) was irradiated with Fe +2 ions at 475 °C for 100, 200, 300 and 400 dpa. Grain coarsening was observed for the samples irradiated for 200–400 dpa resulting in an increase of the average grain size from 152 nm to 620 nm. Growth of submicron grains at higher radiation doses is due to decreased pinning effect imparted by Cr-O rich nanoparticles (NPs) that underwent coarsening via Ostwald ripening. Dislocation density consistently increased with increasing irradiation dose at 300 and 400 dpa. The mean radius of lanthanum-containing nanoclusters (NCs) decreased and their number density increased above 200 dpa, which is likely due to solutes ejection caused by ballistic dissolution and irradiation-enhanced diffusion. Chromium, titanium, oxygen and lanthanum content of nanoclusters irradiated at 200 dpa and higher got reduced by almost half the initial value. The reduction in size of the nanoclusters accompanied with their higher number density and higher dislocation density led to significant radiation hardening with increasing irradiation dose.

FeCrAl alloys with varying compositions and microstructures are under consideration for accident-tolerant fuel cladding, but limited details exist on dislocation loop formation and growth for this class of alloys under neutron irradiation. Four model FeCrAl alloys with chromium contents ranging from 10.01 to 17.51 wt % and aluminum contents of 4.78 to 2.93 wt % were neutron irradiated to doses of 0.3–0.8 displacements per atom (dpa) at temperatures of 335–355 °C. On-zone STEM imaging revealed a mixed population of black dots and larger dislocation loops with either a / 2 〈 111 〉 or a 〈 100 〉 Burgers vectors. Weak composition dependencies were observed and varied depending on whether the defect size, number density, or ratio of defect types was of interest. Results were found to mirror those of previous studies on FeCrAl and FeCr alloys irradiated under similar conditions, although distinct differences exist.

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.

Ferritic-martensitic steels such as T91 and NF616 are candidate materials for several nuclear applications. This study evaluates radiation resistance of T91 and NF616 by examining their microstructural evolutions and hardening after the samples were irradiated in the Advanced Test Reactor to ∼4.3 displacements per atom (dpa) at an as-run temperature of 469 °C. In general, this irradiation did not result in significant difference in the radiation-induced microstructures between the two steels. Compared to NF616, T91 had a higher number density of dislocation loops and a lower level of radiation-induced segregation, together with a slightly higher radiation-hardening. Unlike dislocation loops developed in both steels, radiation-induced cavities were only observed in T91 but remained small with sub-10 nm sizes. Other than the relatively stable M23C6, a new phase (likely Sigma phase) was observed in T91 and radiation-enhanced MX → Z phase transformation was identified in NF616. Laves phase was not observed in the samples.

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 volumetric swelling and hardening effect of irradiated Hastelloy N alloy were investigated in this paper. 7 MeV and 1 MeV Xe ions irradiations were performed at room temperature (RT) with irradiation dose ranging from 0.5 to 27 dpa. The volumetric swelling increases with increasing irradiation dose, and reaches up to 3.2% at 27 dpa. And the irradiation induced lattice expansion is also observed. The irradiation induced hardening initiates at low ion dose (≤1dpa) then saturates with higher ion dose. The irradiation induced volumetric swelling may be ascribed to excess atomic volume of defects. The irradiation induced hardening may be explained by the pinning effect where the defects can act as obstacles for the free movement of dislocation lines. And the evolution of the defects' size and number density could be responsible for the saturation of hardness.

The model for coalescence of randomly distributed immobile pores owing to their growth and impingement, applied by the authors earlier to consideration of the porosity evolution in the high burn-up structure (HBS) at the UO2 fuel pellet periphery (rim zone), was further developed and validated. Predictions of the original model, taking into consideration only binary impingements of growing immobile pores, qualitatively correctly describe the decrease of the pore number density with the increase of the fractional porosity, however notably underestimate the coalescence rate at high burn-ups attained in the outmost region of the rim zone. In order to overcome this discrepancy, the next approximation of the model taking into consideration triple impingements of growing pores was developed. The advanced model provides a reasonable consent with experimental data, thus demonstrating the validity of the proposed pore coarsening mechanism in the HBS.

A nanostructured ferritic alloy (NFA), 14YWT, was produced in the form of thin walled tubing. The stability of the nano-oxides (NOs) was determined under 3.5 MeV Fe +2 irradiations up to a dose of ∼585 dpa at 450 °C. Transmission electron microscopy (TEM) and atom probe tomography (APT) show that severe ion irradiation results in a ∼25% reduction in size between the unirradiated and irradiated case at 270 dpa while no further reduction within the experimental error was seen at higher doses. Conversely, number density increased by ∼30% after irradiation. This ‘inverse coarsening’ can be rationalized by the competition between radiation driven ballistic dissolution and diffusional NO reformation. No significant changes in the composition of the matrix or NOs were observed after irradiation. Modeling the experimental results also indicated a dissolution of the particles.

Thus far, a number of studies have investigated the irradiation evolution of oxide nanoparticles in b.c.c. Fe-Cr based oxide dispersion strengthened (ODS) alloys. But given the inconsistent experimental conditions, results have been widely variable and inconclusive. Crystal structure and chemistry changes differ from experiment to experiment, and the total nanoparticle volume fraction has been observed to both increase and decrease. Furthermore, there has not yet been a comprehensive review of the archival literature. In this paper, we summarize the existing studies on nanoparticle irradiation evolution. We note significant observations with respect to oxide nanoparticle crystallinity, composition, size, and number density. We discuss four possible contributing mechanisms for nanoparticle evolution: ballistic dissolution, Ostwald ripening, irradiation-enhanced diffusion, and homogeneous nucleation. Finally, we propose future directions to achieve a more comprehensive understanding of irradiation effects on oxide nanoparticles in ODS alloys.

FeCrAl alloys are an attractive class of materials for nuclear power applications because of their increased environmental compatibility compared with more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300 and 400 °C have shown post-irradiation microstructures containing dislocation loops and a Cr-rich α′ phase. Although these initial studies established the post-irradiation microstructures, there was little to no focus on understanding the influence of pre-irradiation microstructures on this response. In this study, a well-annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 displacements per atom (dpa) at 382 °C and then the effect of random high-angle grain boundaries on the spatial distribution and size of a〈100〉 dislocation loops, a/2〈111〉 dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with a/2〈111〉 dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and a〈100〉 dislocation loops exhibiting an increased size in the vicinity of the grain boundary. These results suggest the importance of the pre-irradiation microstructure and, specifically, defect sink density spacing to the radiation tolerance of FeCrAl alloys.

FeCrAl alloys have been extensively investigated over the past decade as a candidate material for accident-tolerant fuel cladding in light water reactors. This work completes a first of its kind study where Al and Cr concentrations are varied systematically under neutron irradiation to elucidate the post irradiation microstructural and mechanical response as a direct result of alloying content. Neutron irradiations were performed on alloys with composition Fe-(10-13)Cr-(5-7)Al in wt.% at temperatures of 214, 357, and 557°C to a dose of ~1.8 dpa (displacements per atom). Dislocation loop sizes and dispersion characteristics did not show strong compositional dependence. However, there were noticeable effects of composition on the dispersion of Cr-rich α ′ -precipitates, particularly a decrease in number density with increasing Al content or a decrease in Cr content. The present results confirm previous studies indicating the Cr concentration in these precipitates is lower than that expected in binary FeCr alloys and that Al can act as a destabilizing alloying element for the deleterious α ′ -phase.

The nickel-based Hastelloy N alloy samples were implanted with helium ions at room temperature and then annealed at 600°C (sample S2), 800°C (sample S3) and 900°C (sample S4) for 1 hour, respectively. The effects of post-irradiation isochronous annealing on defects evolution and hardening in the alloy were investigated by transmission electron microscopy (TEM) and nanoindenter. TEM characterization shows that the mean diameter and number density of helium bubbles in sample S2 keep unchanged. It is found that the increments of helium bubbles size in samples S3 and S4 are mainly caused by migration and coalescence (MC) and Ostwald ripening (OR) mechanisms, respectively. Additionally, the MC mechanism can promote the formation of dislocation loops due to matrix atoms diffusion. In sample S4, the recovery effects of defects are striking, resulting in a significant reduction in the number density of dislocation loops. The values of Vac. / He. at different regions have been calculated to explain the size variation of helium bubbles in all the samples. Moreover, nanohardness increments calculated by dispersed barrier hardening (DBH) model agree with those measured by nanoindentation.

High-Cu weld material harvested from an ex service reactor pressure vessel (RPV) steel from Unit 1 of the decommissioned Zion Nuclear Generating Station has been characterized using atom probe tomography. Samples taken from 4 different positions through the thicknesses of the pressure vessel wall from the water-side to the air-side were characterized, along with an unirradiated baseline material. In the baseline material, no precipitates were found and the Cu was observed to be fully in solid solution; however, scanning transmission electron microscopy combined with energy dispersive spectroscopy (STEM-EDS) revealed the presence of ϵ -Cu that form during processing of the material and results in the concomitant decrease of matrix Cu. Following irradiation, a high number density of nano-scale Cu-rich precipitates (CRPs) were observed, uniformly distributed throughout the matrix. The Cu content within the CRPs was found to be ∼ 30-35 at.% regardless of location in the wall. No statistically significant variation in the compositions, mean radius, number density, or volume fraction as a function of location within the wall was observed. The measured matrix Cu level excluding CRPs contribution was found to be ∼ 90 appm higher than the solubility limit suggesting that further nucleation and growth of the CRPs under continued operations would have occurred. These results clearly demonstrate that the neutron energy attenuation has no significant effect on the precipitation kinetics of CRPs regardless of location in the wall in high-Cu RPV steels under irradiation. Notice: This manuscript has been authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ).

In order to understand the effect of stacking fault energy on microstructural evolution in concentrated solid solution alloys, Kr2+ ion irradiation was performed for FCC-type CoCrFeNiMnx alloys. During Kr2+ ion irradiation at 500 °C the formation of black dots, self-interstitial atom faulted loops and perfect loops were observed, but no observable voids were found in all the alloys. The comparison of microstructural evolution among CoCrFeNiMnx (x = 0.7, 1, 1.3) alloys revealed less faulted loop number density and size in CoCrFeNiMn1.3. The stacking fault energy of each alloy was investigated using the deformed CoCrFeNiMnx alloys, and it was revealed that CoCrFeNiMn1.3 had the highest stacking fault energy, which is higher value than that of 316SS. From these results, it is suggested that FeCrNiMn-based concentrated solid solution alloys could be well-designed by controlling the stacking fault energy with optimized Mn concentration as one of high radiation resistance structure materials.

In this study, the microstructural changes induced by doping of Nb in Zr were investigated by the combined utilization of electron backscatter diffraction and electron transmission microscopy techniques, followed by the correlated hardening mechanism being elucidated based on the obtained microstructural parameters. Microstructural characterization results revealed that microstructural changes caused by doping of Nb in Zr were mainly embodied via two aspects: reducing the matrix α-Zr grain size and increasing the amount of β-Nb particles. β-phase stabilizing effect, dragging effect and pinning effect introduced and enhanced by Nb addition, worked together to significantly reduce the grain size in Zr–Nb alloys. β-Nb particles were firstly observed in Zr0.5Nb specimen with the fairly low number density of ∼2.0 × 10 18 /m 3 , then this value explosively increased to ∼3.3 × 10 20 /m 3 for Zr2Nb specimen. In addition, hardness was increased with an increase in the Nb content. The hardening contributions from solid solution hardening, grain boundary hardening and precipitation hardening were quantitatively estimated as per the obtained microstructural parameters. Results inferred that solid solution hardening contributed the majority when the Nb atoms were solid dissolved (≤0.5 wt%), whereas the precipitation hardening surpassed any other factors when the β-Nb particles were steadily precipitated (≥1 wt%).

Nanostructure evolution during ion irradiation of two thermally aged binary Fee22Cr alloys has been investigated using atom probe tomography. Specimens aged at 500 °C for 50 and 200 h were irradiated by 5.6 MeV Fe ions at room temperature up to fluences of 0.3 × 1015 ions/cm2 and 1 × 1015 ions/cm2. The effect of irradiation on the material nanostructure was examined at a depth of 1 ?m from the irradiated surface. The analysis of Cr radial concentration functions reveals that dense ??-phase precipitates in the 200 h aged alloy become diffuse and thereby larger when subjected to irradiation. On the other hand, less Cr-enriched precipitates in the alloy aged for 50 h are less affected. The CreCr pair correlation function analysis shows that matrix inhomogeneity decreases under irradiation. Irradiation leads to a decrease in the number density of diffuse clusters, whereas in the case of well-developed precipitates it remains unchanged.

Two 9Cr-ODS steels with the same nominal composition were consolidated by hot isostatic pressing (HIP, named COS-1) and spark plasma sintering (SPS, named COS-2). Helium ions were implanted into COS-1, COS-2 and non-ODS Eurofer 97 steels up at 673 K. Microstructures before and after helium ion implantations were carefully characterized. The results show a bimodal grain size distribution in COS-2 and a more uniform grain size distribution in COS-1. Nanoscale clusters of GP-zone type Y–Ti–O and Y 2 Ti 2 O 7 pyrochlore as well as large spinel Mn(Ti)Cr 2 O 4 particles are all observed in the two ODS steels. The Y–Ti-enriched nano-oxides in COS-1 exhibit higher number density and smaller size than in COS-2. The Y–Ti-enriched nano-oxides in fine grains of COS-2 show higher number density and smaller size than that in coarse grains of COS-2. Nano-oxides effectively trap helium atoms and lead to the formation of high density and ultra-fine helium bubbles.

A lanthana-containing nanostructured ferritic steel (NFS) was processed via mechanical alloying (MA) of Fe-14Cr-1Ti-0.3Mo-0.5La2O3 (wt.%) and consolidated via spark plasma sintering (SPS). In order to study the consolidation behavior via SPS, sintering temperature and dwell time were correlated with microstructure, density, microhardness and shear yield strength of the sintered specimens. A bimodal grain size distribution including both micron-sized and nano-sized grains was observed in the microstructure of specimens sintered at 850, 950 and1050 °C for 45 min. Significant densification occurred at temperatures greater than 950 °C with a relative density higher than 98%. A variety of nanoparticles, some enriched in Fe and Cr oxides and copious nanoparticles smaller than 10 nm with faceted morphology and enriched in La and Ti oxides were observed. After SPS at 950 °C, the number density of Cr–Ti–La–O-enriched nanoclusters with an average radius of 1.5 nm was estimated to be 1.2 × 1024 m−3. The La + Ti:O ratio was close to 1 after SPS at 950 and 1050 °C; however, the number density of nanoclusters decreased at 1050 °C. With SPS above 950 °C, the density improved but the microhardness and shear yield strength decreased due to partial coarsening of the grains and nanoparticles.

Cold-worked 316 stainless steel specimens irradiated to 74 dpa in a pressurized water reactor (PWR) were analyzed by atom probe tomography (APT) to extend knowledge of solute clusters and segregation at higher doses. The analyses confirmed that those clusters mainly enriched in Ni–Si or Ni–Si–Mn were formed at high number density. The clusters were divided into three types based on their size and Mn content; small Ni–Si clusters (3–4 nm in diameter), and large Ni–Si and Ni–Si–Mn clusters (8–10 nm in diameter). The total cluster number density was 7.7 × 1023 m−3. The fraction of large clusters was almost 1/10 of the total density. The average composition (in at%) for small clusters was: Fe, 54; Cr, 12; Mn, 1; Ni, 22; Si, 11; Mo, 1, and for large clusters it was: Fe, 44; Cr, 9; Mn, 2; Ni, 29; Si, 14; Mo,1. It was likely that some of the Ni–Si clusters correspond to γ′ phase precipitates while the Ni–Si–Mn clusters were precursors of G phase precipitates. The APT analyses at grain boundaries confirmed enrichment of Ni, Si, P and Cu and depletion of Fe, Cr, Mo and Mn. The segregation behavior was consistent with previous knowledge of radiation induced segregation.

Nanostructured ferritic alloys (NFAs) are new generation materials for use in high temperature energy systems, such as nuclear fission or fusion reactors. However, joining these materials is a concern, as their unique microstructure is destroyed by traditional liquid-state welding methods. The microstructural evolution of a friction stir welded 14YWT NFA was investigated by atom probe tomography, before and after a post-weld heat treatment (PWHT) at 1123K. The particle size, number density, elemental composition, and morphology of the titanium-yttrium-oxygen-enriched nanoclusters (NCs) in the stir and thermally-affected zones were studied and compared with the base metal. No statistical difference in the size of the NCs was observed in any of these conditions. After the PWHT, increases in the number density and the oxygen enrichment in the NCs were observed. Therefore, these new results provide additional supporting evidence that friction stir welding appears to be a viable joining technique for NFAs, as the microstructural parameters of the NCs are not strongly affected, in contrast to traditional welding techniques.