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In this study, a numerical methodology is developed to monitor stress distributions and an experimental technique is investigated to measure stresses on metal embedded polymer matrix fiber-reinforced composite materials using X-ray diffraction (XRD) method. The numerical method successfully predicts the stress distributions using only a few data points with an average accuracies of 11.6% and 11.8%, respectively for tapered and open hole tensile sample geometries. Experimental part of this study investigates different methodologies to measure stresses on surfaces of polymer composites materials of thin walled structures by XRD. For this reason, metallic materials were integrated onto the surface and near-surface of polymer matrix composites in three different configurations as in the forms of metal foil, metal powder, and metal wire mesh. In-situ experiments of measured and applied stresses revealed metal powder embedded composite materials as a promising material configuration for reliable monitoring of stresses.

The nanostructured ferritic alloy 14YWT is a promising candidate for in-core use in generation IV nuclear reactors, due to a dense dispersion of insoluble, ultrafine-scale Y-Ti-O nano-oxides, which provide a high degree of irradiation tolerance and thermal stability. This study investigates the effects of heat treatment on the workability of 14YWT, along with the effect of processing history on abnormal grain structures and radial microstructural uniformity. In this study, a 14YWT rod consolidated at 850 °C was heat treated in argon at 1100 °C, 1150 °C, 1200 °C, or 1250 °C for 1 or 8 hours and changes in mechanical properties and microstructure were examined using microhardness and electron backscattered diffraction (EBSD). Two distinct types of large abnormal grains were observed, each with unique processing origins, including one with high and the other with low strain energy. Consolidation via direct extrusion resulted in radial microstructural gradients, where the center of the rod was softer with larger grain sizes and lower strain energies. These gradients persisted and intensified throughout heat treatment. Based upon this work, the recommended heat treatment for increased workability with minimal microstructural change is 1150 °C for 1 hours.

The effects of 5 MeV Fe2+ ion irradiation at 300°C on the microstructure evolution and deformation behavior of a FeCrAl C26M alloy are presented. It has been found that dislocation loop density increases an order of magnitude from 1 dpa to 16 dpa irradiations, whereas, the dislocation loop size saturates with increasing damage. Micropillars, 600 nm in diameter and 1.3 µm in height, were fabricated and compressed inside grains with , and crystallographic orientations, respectively. {112} has been identified as the primary slip system in both unirradiated and irradiated alloys. The increase in yield stress after irradiation is observed with measurable variation along and vs. along . By applying the Orowan dispersed barrier model, the increase of yield stress is found mainly due to the slip resistance of radiation generated defect loops. Detailed transmission electron microscopy (TEM) studies were performed to quantify the Burgers vector and the distribution of irradiation induced dislocations at elevated strains. It is revealed that localized shear instability is caused by avalanche slip events of ½ dislocations gliding out of tested pillars. Simultaneously, a large number of sessile/immobile dislocations formed in the vicinity of slip band, leading to the hardening at elevated strains.

Ferritic/martensitic steels are one of the best candidates for structural materials for high dose applications in next generation nuclear reactors. The composition of structural materials must be optimized for reliable service during irradiation. This study reports the effect of interstitial nitrogen on radiation response in 12Cr ferritic/martensitic HT9 steels having a controlled nitrogen concentration. Results show that a high amount of ‘free’ nitrogen in the matrix stabilizes the interstitial clusters which leads to (i) larger loop sizes (ii) lower loop density and (iii) slightly reduced radiation induced hardening. It also affects diffusion mechanism of Ni and formation of Ni/Si-rich precipitates.

Inconel alloy 718 is a high-strength and corrosion resistant alloy that is commonly used as a beamline vacuum window. The accumulation of irradiation-induced damage substantially decreases the window’s service lifetime, and replacing it engenders significant beamline downtime. With this application in mind, herein we examine whether post-irradiation annealing can alleviate irradiation-induced damage of Inconel alloy 718. Inconel alloy 718 was received in a solution annealed state. We then irradiated samples using two different modalities (1.5 MeV H+ and 5 MeV Ni2+) at three representative temperatures for beamline windows (room temperature, 100 °C, and 200 °C), followed by annealing at temperatures viable for in-situ annealing processes (no anneal, 300 °C, and 500 °C). Using nanoindentation, we determined that irradiation-induced hardening occurs but is largely mitigated by post-irradiation annealing. Overall, our results suggest that in-situ annealing of radiation damage in Inconel alloy 718 vacuum windows appears feasible, which could potentially decrease beam downtime and maintenance costs.

Nanostructured ferritic alloys (NFAs) having sub-micron grain size with a high density of nano-oxides (NOs) (size of ∼2–3 nm) are one of the best candidates for structural components in Generation IV nuclear systems. In this study, 14YWT NFA cladding tubes were irradiated in BOR60 reactor up to 7 dpa at 360–370 °C. Detailed microstructural analysis has been conducted using bright field transmission electron microscopy, bright field scanning transmission electron microscopy, energy filtered transmission electron microscopy, energy dispersive X-ray spectroscopy, electron energy loss spectroscopy and transmission Kikuchi diffraction techniques. This revealed cavities, and type dislocation loops, and α′ precipitates forming after irradiation with relationships between cavities and NOs, and α′ precipitates and NOs. Cavities mostly form on the NOs; whereas, α′ precipitates form between the NOs where the point defect concentration is high. Moreover, α′ precipitates are distributed homogenously on and around the dislocation loops which is consistent with the observation that there is no Cr segregation on dislocation loops. Grain boundaries were found to be mostly depleted in Cr; however, the characteristics of each grain boundary determines the Cr behavior and the α′ denuded zone around the grain boundaries. Mechanical properties of the irradiated tubes have been determined by using both low force and high force nanoindentation techniques, resulting in 1.03 ± 0.33 GPa and 0.82 ± 0.20 GPa hardening, respectively. Dispersed barrier hardening calculations and nanoindentation measurements are in good agreement. In this study, 14YWT NFA has been systematically studied after neutron irradiation to better understand its superior performance: low α′ concentration, low swelling and low radiation-induced hardening.

Accelerator-based ion irradiation is commonly used to simulate neutron damage, in lieu of neutron irradiation due to limited availability of fast flux facilities and little to no activation of the samples. Neutron atypical effects, however, must be recognized and their impact minimized in order to achieve the most accurate microstructural evolution under ion bombardment. Mechanical property changes, which arise from the synergy of numerous radiation-induced changes, are especially susceptible to these neutron atypical effects. Of these effects, a recently studied neutron atypical effect, arising from Coulomb-drag of vacuum contaminants and subsequent compositional alteration, was shown to substantially suppress swelling. Through the use of an ion beam filtering system, the resulting void swelling was shown to closely match neutron irradiation data. In this study, the impact of compositional modification via Coulomb-drag on the mechanical property changes is examined through the use of nanoindentation on a ferritic/martensitic alloy, HT9, irradiated with and without the use of the ion beam filtering system.

Nanocrystalline metals are often postulated as irradiation tolerant materials due to higher grain boundary densities. The efficiency of these materials in mitigating irradiation damage is still under investigation. Here, we present an in-situ transmission electron microscopy with ion irradiation study on equiaxed 35 nm grained tungsten (NCW-35 nm) and compare its radiation tolerance, in terms of dislocation loop damage, to several other grades of tungsten with different grain sizes at two temperatures (RT and 1073 K). The NCW-35 nm was shown to possess significant higher radiation tolerance in terms of loop damage. As demonstrated by Kinetic Monte Carlo simulations, at least part of the higher radiation tolerance of the small grains is due to higher interstitial storage (at the grain boundaries) and defect recombination (in the grain interiors) in the small grain material. In addition, experimental observations reveal rapid and efficient dislocation loop absorption by the grain boundaries and this is considered the dominant factor for mass transport to the boundaries during irradiation, enabling the remarkable radiation tolerance of 35 nm grained tungsten. This study demonstrates the possibility of attaining high radiation tolerant materials, in terms of dislocation loop damage, by minimizing grain sizes in the nanocrystalline regime.

The effects of 5 MeV Fe2+ ion irradiation at 300 °C on the microstructure evolution and deformation behavior of a FeCrAl C26M alloy are presented. It has been found that dislocation loop density increases an order of magnitude from 1 dpa to 16 dpa irradiations, whereas, the dislocation loop size saturates with increasing damage. Micropillars, 600 nm in diameter and 1.3 um in height, were fabricated and compressed inside grains with , and crystallographic orientations, respectively. {112} has been identified as the primary slip system. The irradiation-induced hardening generally follows the prediction of Orowan dispersed barrier hardening equation, even though the exact increasement of yield stress after irradiation is presented heterogeneously across three orientations. Detailed transmission electron microscopy (TEM) studies were performed to quantify the Burgers vector and the distribution of irradiation induced dislocations. It is revealed that localized shear instability is caused by avalanche slip events of ½ dislocations gliding out of tested pillars. Simultaneously, a large number of sessile/immobile dislocations formed in the vicinity of slip band, leading to the hardening at elevated strains.

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al2O3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe2+ ion irradiation up to ∼16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two-beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α′ precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6–2.0 GPa hardening for both tubes between ∼3.4 dpa and ∼16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

Displacement damage, through heavy ion irradiation was studied on two tungsten grades (coarse grained tungsten (CGW) and nanocrystalline and ultrafine grained tungsten (NCW)) using different displacement per atom rates and different irradiation temperatures (RT and 1050 K). Percentage of and type loops at the irradiation conditions was determined. Irradiation damage in the microstructure was quantified using average loop areas and densities (method A) and loop areal fraction in the grain matrices under 2-beam diffraction conditions (method B). Average values of and loops were calculated from method A. Loop coalescence was shown to occur for CGW at 0.25 dpa. Using both methods of quantifying microstructural damage, no effect of dpa rate was observed and damage in CGW was shown to be the same at RT and 1050 K. Swelling from voids observed at 1050 K was quantified. The loop damage in NCW was compared to CGW at the same diffraction and imaging conditions. NCW was shown to possess enhanced irradiation resistance at RT regarding loop damage and higher swelling resistance at 1050 K compared to CGW. For irradiation at 1050 K, the NCW was shown to have similar defect densities to the CGW which is attributed to higher surface effects in the CGW, vacancy loop growth to voids and a better sink efficiency in the CGW deduced from the vacancy distribution profiles from Kinetic Monte Carlo simulations. Loop density and swelling was shown to have similar values in grains sizes that range from 80 to 600 nm. No loop or void denuded zones occurred at any of the irradiation conditions. This work has a collection of experiments and conclusions that are of vital importance to materials and nuclear energy communities.

In-situ neutron diffraction experiments were performed on room temperature compressed 14YWT nanostructured ferritic alloys at 1100 °C and 1150 °C to understand their thermally activated static recrystallization mechanisms. Existence of a high density of Y-Ti-O rich nano-oxides ( and {112} texture components during annealing, in contrast to the conventional recrystallization textures in body centered cubic alloys. Furthermore, nano-oxide size, shape, density and distribution are considerably different in unrecrystallized and abnormally grown grains. Transmission electron microscopy analysis shows that oxide particles having a size between 5 and 30 nm play a critical role for recrystallization mechanisms in 14YWT nanostructured ferritic alloys.

Dislocation loop rafting and dislocation decoration have been previously observed in neutron and heavy ion irradiated materials. Understanding the fundamental aspects of these phenomena assist in evaluating irradiation damage of nuclear materials. Multiple different mechanisms have been suggested to explain loop rafting. Here, we performed a detailed transmission electron microscopy study on dislocation loop rafts in heavy ion irradiated tungsten. Different imaging conditions showed that the rafts are of Burgers vector type and specifically the same Burgers vector variant ( 1 ¯ 1>) in the particular grain analyzed. Some rafts were associated with dislocation lines while some form as a result of alignment of dislocation loops. They were shown to form at both room temperature (RT) and high temperature with stronger rafts forming at RT. These observations confirm the mechanism previously suggested by Wen et al. which explains raft formation due to loop glide, cluster-cluster and grown-in dislocation-cluster interaction with subsequent Burgers vector rotation. Similar irradiation studies on nanocrystalline tungsten showed that these materials are more resistant to raft formation at RT irradiations.

Although accelerator-based ion irradiation has been widely accepted to simulate neutron damage, neutron-atypical features need to be carefully investigated. In this study, we have shown that Coulomb force drag by ion beams can introduce significant amounts of carbon, nitrogen, and oxygen into target materials even under ultra-high vacuum conditions. The resulting compositional and microstructural changes dramatically suppress void swelling. By applying a beam-filtering technique, introduction of vacuum contaminants is greatly minimized and the true swelling resistance of the alloys is revealed and matches neutron behavior closely. These findings are a significant step toward developing standardized procedures for emulating neutron damage.

In this research, innovative thermal spray deposition (Process I) and conventional hot extrusion processing (Process II) methods have been used to produce thin walled tubing (∼0.5 mm wall thickness) out of 14YWT, a nanostructured ferritic alloy. The effects of processing methods on the microstructure, mechanical properties and irradiation response have been investigated by using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and, micro- and nano-hardness techniques. It has been found that these two processes have a significant effect on the microstructure and mechanical properties of the as-fabricated 14YWT tubes. Even though both processing methods yield the formation of various size Y-Ti-O particles, the conventional hot extrusion method results in a microstructure with smaller, homogenously distributed nano-oxides (NOs, Y-Ti-O particles

EK-181 is a low-activation ferritic/martensitic steel that is an attractive candidate for in-core component materials for both fast reactors and fusion reactors. To assess the effect of microstructural engineering on radiation response, two variants of EK-181 were studied: one in an annealed condition and the other subject to severe plastic deformation. These specimens were irradiated with 3.5 MeV Fe self-ions up to 400 peak displacements per atom (dpa) at temperatures ranging from 400 °C to 500 °C. The deformation did not suppress swelling over the whole irradiated region. Instead, deformed samples showed higher swelling in the near-surface region. Void swelling was found to be correlated with grain boundary instability. Significant grain growth occurred when steady-state void growth started.

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.

The phase properties and deformation behavior of the δ –ferrite and γ –austenite phases of CF–3 and CF–8 cast duplex stainless steels were characterized by nanoindentation and microstructure-based finite element method (FEM) models. The elastic modulus of each phase was evaluated and the results indicate that the mean elastic modulus of the δ –ferrite phase is greater than that of the γ –austenite phase, and the mean nanoindentation hardness values of each phase are approximately the same. The elastic FEM model results illustrate that greater von Mises stresses are located within the δ –ferrite phase, while greater von Mises strains are located in the γ –austenite phase in response to elastic deformation. The elastic moduli calculated by FEM agree closely with those measured by tensile testing. The plastically deformed specimens exhibit an increase in misorientation, deformed grains, and subgrain structure formation as measured by electron backscatter diffraction (EBSD).

Ferritic/martensitic alloys are required for advanced reactor components to survive 500–600 neutron-induced dpa. Ion-induced void swelling of ferritic/martensitic alloy T91 in the quenched and tempered condition has been studied using a defocused, non-rastered 3.5 MeV Fe-ion beam at 475 °C to produce damage levels up to 1000 peak displacements per atom (dpa). The high peak damage level of 1000 dpa is required to reach 500–600 dpa level due to injected interstitial suppression of void nucleation in the peak dpa region, requiring data extraction closer to the surface at lower dpa levels. At a relatively low peak damage level of 250 dpa, voids began to develop, appearing first in the near-surface region. With increasing ion fluence, swelling was observed deeper in the specimen, but remained completely suppressed in the back half of the ion range, even at 1000 peak dpa. The local differences in dpa rate in the front half of the ion range induce an “internal temperature shift” that strongly influences the onset of swelling, with shorter transient regimes resulting from lower dpa rates, in agreement not only with observations in neutron irradiation studies but also in various ion irradiations. Swelling was accompanied by radiation-induced precipitation of Cu-rich and Si, Ni, Mn-rich phases were observed by atom probe tomography, indicating concurrent microchemical evolution was in progress. In comparison to other ferritic/martensitic alloys during ion irradiation, T91 exhibits good swelling resistance with a swelling incubation period of about 400 local dpa.

Nanostructured ferritic alloys are considered as candidates for structural components in advanced nuclear reactors due to a high density of nano-oxides (NOs) and ultrafine grain sizes. However, bimodal grain size distribution results in inhomogeneous NO distribution, or vice versa. Here, we report that density of NOs in small grains (2 µm) before and after irradiation. After 6 dpa neutron irradiation at 385–430 °C, α′ precipitation has been observed in these alloys; however, their size and number densities vary considerably in small and large grains. In this study, we have investigated the precipitation kinetics of α′ particles based on the sink density, using both transmission electron microscopy and kinetic Monte Carlo simulations. It has been found that in the presence of a low sink density, α′ particles form and grow faster due to the existence of a larger defect density in the matrix. On the other hand, while α′ particles form far away from the sink interface when the sink size is small, Cr starts to segregate at the sink interface with the increase in the sink size. Additionally, grain boundary characteristics are found to determine the radiation-induced segregation of Cr.

© 2020 Elsevier B.V.Laser powder bed additively manufactured Grade 91 composition steel was investigated in comparison to wrought Grade 91 steel in terms of microstructure and mechanical properties. As-deposited additively manufactured Grade 91 steel had a microstructure of lower bainitic regions surrounded by martensite. This is significantly different from the typical tempered martensitic microstructure of conventionally produced Grade 91 steel. The as-deposited additively manufactured material had excellent tensile mechanical properties with greater strength than the wrought material at room temperature, 300 and 600°C showing excellent promise for nuclear applications. Retention of strength at 300 and 600°C for the as-deposited additively manufactured material was attributed to transitional carbides in the lower bainitic regions. The additively manufactured material was also investigated in the tempered as well as normalized and tempered conditions, each showing decreased strength at elevated temperature than the as-deposited material.

Nanostructured ferritic alloys (NFAs) are considered as candidates for structural components in advanced nuclear reactors due to their excellent radiation resistance as a result of a high density of nano-oxides (NOs) in the microstructure. Therefore, gaining an understanding on the stability of NOs under irradiation is crucial. In this study, we have investigated the evolution of defects and NOs in 14YWT NFAs under in-situ Kr ion irradiation at room temperature (RT) and 450 °C up to 10 dpa. It has been found that irradiations at 450 °C do not create any changes in the NOs, similar to the bulk irradiations. On the other hand, elemental mapping indicates that NOs dissolve mostly after 10 dpa irradiations at RT. Thus, while defects are both annihilated and pinned by NOs at low doses (before the dissolution of NOs), glissile loops start to escape to the foil surface at high doses (after the dissolution of NOs), justifying the significantly low fraction of loops compared to the literature values. High resolution transmission electron microscopy analysis has shown that the NOs are mostly coherent Y2Ti2O7 particles with pyrochlore crystal structure after both RT and 450 °C irradiations, similar to those observed before irradiation.

In this study, the microstructure of a 12Cr ferritic-martensitic oxide-dispersion-strengthened (ODS) alloy is studied before and after Fe ion irradiation up to 200 peak displacements per atom (dpa). Irradiation temperature ranges from 325 to 625 °C. Before irradiation, both coherent and incoherent dispersoids exist in the matrix. In response to irradiation, the mean sizes of dispersoids in both the ferrite and tempered martensite phases change to equilibrium values that increase with irradiation temperature. The evolution of dispersoids under irradiation is explained by a competition between athermal-radiation-driven shrinkage and thermal-diffusion-driven growth, with interface coherency affecting the growth mechanism. However, each coherency type exhibits different evolution behavior under irradiation. Coherent dispersoids, regardless of their initial size, change toward an equilibrium size at each temperature tested. On the other hand, incoherent dispersoids are destroyed at lower test temperatures but survive while shrinking in size at higher temperatures. This difference in behavior can be explained by the lower interfacial energy of coherent dispersoids in comparison with incoherent dispersoids. This study sheds light on the roles of interface configurations in maintaining dispersoid integrity under irradiation.

Texture and microstructure of tubes and plates fabricated from a nanostructured ferritic alloy (14YWT), produced either by spray forming followed by hydrostatic extrusion (Process I) or hot extrusion and cross-rolling a plate followed by hydrostatic tube extrusion (Process II) have been characterized in terms of their effects on texture and grain boundary character. Hydrostatic extrusion results in a combination of plane strain and shear deformations which generate low intensity α- and γ-fiber components of {001} and {111} together with a weak ζ-fiber component of {011} and {011} . In contrast, multi-step plane strain deformation by hot extrusion and cross-rolling of the plate leads to a strong texture component of {001} together with a weaker {111} component. Although the total strains are similar, shear dominated deformation leads to much lower texture indexes compared to plane strain deformations. Further, the texture intensity decreases after hydrostatic extrusion of the alloy plate formed by plane strain deformation, due to a lower number of activated slip systems during shear dominated deformation. Notably, hot extruded and cross-rolled plate subjected to plane strain deformation to ~50% engineering strain creates only a modest population of low angle grain boundaries, compared to the much larger population observed following the combination of plane strain and shear deformation of ~44% engineering strain resulting from subsequent hydrostatic extrusion.

The microstructural changes and mechanical response of an HT-9 sample shock loaded to a peak pressure of 11 GPa have been investigated by TEM, XRD, microhardness and EBSD techniques. Dislocation densities obtained by both direct measurements (via TEM) and indirect calculations (by XRD and hardness) indicate that shock loading results in ~2–3 fold increase in dislocation density. TEM analyses show that the shape, and density of the dislocations change after shock loading. In addition, shock loading causes local plastic deformation of the continuous parallel lath structure in some regions, together with an overall decrease in the aspect ratio of laths due to local plastic deformation and lath fragmentation. As a result of XRD analyses, the fraction of edge dislocations is determined to increase by ~24% after shock loading. Furthermore, hardness increases by ~40 HV after shock loading due to the increased dislocation density. EBSD analyses show that the fraction of CSL boundaries decreases by ~5–10% as a result of shock loading.

The changes due to pulsed plasma flow irradiation on the near-surface microstructure and mechanical properties of the high-chromium, ferritic–martensitic steel EK-181 (Fe16Cr12W2VTaB) have been quantified. Irradiation of EK-181 in this manner produces a microstructural gradient near the material surface, with a two dimensional nanostructured cellular surface. The microstructure and mechanical properties of the modified layer are independent of the initial microstructure and phase composition, and are strongly defined solely by parameters of the plasma flow. High thermal stability of the pulsed plasma-modified layer was explicitly demonstrated.

A dual-phase 12Cr oxide-dispersion-strengthened (ODS) alloy, with improved corrosion and oxidation resistance exhibits promising void swelling resistance and microstructural stability under Fe2+ ion irradiation to 800 dpa at 475 °C. Dispersoids were originally present in both ferrite and tempered martensite grains, with the latter having a wider range of dispersoid sizes. In both phases dispersoids >10 nm in diameter are incoherent with the matrix, while smaller dispersoids are coherent. During irradiation the larger incoherent dispersoids shrank and disappeared. Beyond 60 dpa dispersoids in both phases approached a near-identical equilibrium size of ∼2–2.5 nm, which appears to be rather independent of local displacement rate. Grain morphology was found to be stable under irradiation. Compared to other ferritic-martensitic alloys, the ion-induced swelling of this alloy is quite low, arising from swelling resistance associated with both tempered martensite and dispersoids in both phases. Swelling in tempered martensite is an order of magnitude less than in the ferrite phase.

Ion beam scanning or “rastering” is a technique that is frequently used to uniformly cover a larger specimen area during ion irradiation. In this study, we addressed the effects of rastered and defocused beams, using 3.5 MeV iron ions to irradiate pure iron at 450 °C to peak doses of 50 and 150 dpa. We focused on a frequency range relevant to pulsed fusion devices and show its importance to ion irradiation experiments used for simulating neutron damage. The beam was scanned at 15.6, 1.94, and 0.244 Hz and the resulting microstructure was compared with that produced by a non-rastered, defocused beam. At 150 dpa, the defocused beam case resulted in the highest observed void swelling of ∼12% at a depth of ∼700 nm, a depth short of the peak dose position at 1000 nm. The swelling at the peak dose position was significantly reduced by the defect imbalance phenomenon. A maximum swelling rate of ∼0.12%/dpa was measured in this specimen at a depth of 600 nm below the ion-incident surface. Rastering led to much lower swelling levels achieved at significantly lower swelling rates, with the greatest rate of decrease occurring below ∼1 Hz. Furthermore, the impact of the defect imbalance arising from interstitial injection and spatial distribution difference of initial interstitial and vacancy defects was strongly pronounced in the non-rastered case with a lesser effect observed with decreasing raster frequency.

Micron size barium hexaferrite (BaFe 12 O 19 ) platelets were prepared by the molten-salt synthesis method in various weight proportions of NaCl–KCl salt mixtures as a liquid medium. The effect of molten salt composition— x wt% NaCl and (100− x ) wt% KCl—on the amount of barium hexaferrite phase formation, as well as, on the morphology and magnetic properties of the final products are discussed. Inductively coupled plasma-mass spectroscopy (ICP-MS) was used to determine the solubility of the starting materials in the salts in order to understand the formation mechanism of barium hexaferrite. X-ray diffraction analysis (XRD), scanning electron microscopy (SEM) and vibrating sample magnetometer (VSM) were used to identify the characteristics of the synthesized barium hexaferrite platelets. ICP-MS analysis showed that the Fe 2 O 3 solubility is negligible, however, BaCO 3 has very high solubility in both molten KCl and NaCl. Quantitative XRD and SEM results showed that molten salt containing 100 wt% NaCl at 900 °C (for 2 h) resulted in the highest production of barium hexaferrite. Further, SEM results showed that KCl-rich molten salt led to the formation of sharper faceted platelet morphology, whereas NaCl-rich ones resulted in more round platelets. Data from magnetic measurements showed that as the content of NaCl in the molten salt increases, hysteresis losses became higher. This is a characteristic of the achievement of a harder magnetite behavior in the synthesized barium hexaferrite ceramics. The composition of the KCl/NaCl molten salts was shown to play an important role on the extent of barium hexaferrite formation, resulting platelet morphology and the magnetic properties.

Fabrication methods are being developed and tested for producing fuel clad tubing of the advanced ODS 14YWT and FCRD-NFA1 ferritic alloys. Three fabrication methods were based on plastically deforming a machined thick-wall tube sample of the ODS alloys by pilgering, hydrostatic extrusion or drawing to decrease the outer diameter and wall thickness and increase the length of the final tube. The fourth fabrication method consisted of the additive manufacturing approach involving solid-state spray deposition (SSSD) of ball milled and annealed powder of 14YWT for producing thin-wall tubes. Of the four fabrication methods, two methods were successful at producing tubing for further characterization: production of tubing by high-velocity oxy-fuel spray forming and production of tubing using high-temperature hydrostatic extrusion. The characterization described shows through neutron diffraction the texture produced during extrusion while maintaining the beneficial oxide dispersion. In this research, the parameters for innovative thermal spray deposition and hot extrusion processing methods have been developed to produce the final nanostructured ferritic alloy (NFA) tubes having approximately 0.5 mm wall thickness. Effect of different processing routes on texture and grain boundary characteristics has been investigated. It was found that hydrostatic extrusion results in combination of plane strain and shear deformations which generatemore » rolling textures of α- and γ-fibers on {001} and {111} together with a shear texture of ζ-fiber on {011} and {011} . On the other hand, multi-step plane strain deformation in cross directions leads to a strong rolling textures of θ- and e-fiber on {001} together with weak γ-fiber on {111} . Even though the amount of the equivalent strain is similar, shear deformation leads to much lower texture indexes compared to the plane strain deformations. Moreover, while 50% of hot rolling brings about a large number of high-angle grain boundaries (HAB), 44% of shear deformation results in large amount of low-angle boundaries (LAB) showing the incomplete recrystallization.« less

Fabrication of thin walled tubing of FeCrAl alloys is critical to its success as a candidate enhanced accident tolerant fuel cladding material. Alloys that are being investigated are Generation I and Generation II FeCrAl alloys produced at ORNL and an ODS FeCrAl alloy, MA-956 produced by Special Metals. Gen I and Gen II FeCrAl alloys were provided by ORNL and MA-956 was provided by LANL (initially produced by Special Metals). Three tube development efforts were undertaken. ORNL led the FeCrAl Gen I and Gen II alloy development and tube processing studies through drawing tubes at Rhenium Corporation. LANL received alloys from ORNL and led tube processing studies through drawing tubes at Century Tubing. PNNL led the development of tube processing studies on MA-956 through pilger processing working with Sandvik Corporation. A summary of the recent progress on tube development is provided in the following report and a separate ORNL report: ORNL/TM-2015/478, “Development and Quality Assessments of Commercial Heat Production of ATF FeCrAl Tubes”.

The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commercially available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.