Candidate Structural Materials Research Articles

Atomistic simulation of chemical short-range order on the irradiation resistance of HfNbTaTiZr high entropy alloy

High entropy alloys (HEAs) have been considered as one of the potential structural material candidates for fourth-generation nuclear reactors and fusion reactors due to their excellent irradiation resistance. Current studies have shown that the chemical short-range order (CSRO) usually exists in HEAs, which has a significant effect on the mechanical properties and irradiation resistance of HEAs. Refractory high entropy alloys (RHEAs), as a new class of HEAs have better mechanical properties at high temperatures than face-centered cubic (FCC) HEAs, and therefore have better prospects of application in the nuclear field. In this study, CSRO and its effect on the irradiation resistance of HfNbTaTiZr are analyzed via molecular dynamics (MD) and Monte Carlo (MC). The primary cascade simulations, multi-cascade simulations and surface bombardment simulations are carried out to simulate the generation and accumulation of irradiation damage. The results of the primary cascade simulations and surface bombardment simulations of CSRO models show that the presence of CSRO induces cascade splitting into subcascades. The presence of subcascades reduces the thermal peak enhancement effect and thus lowers the recombination rate of Frenkel pairs (FPs) in the damage zone when FPs concentrations are low. However, the creation of subcascades increases the size of the damage zone caused by the cascade. Thus, when the concentrations of FPs are high, the larger area of damage zone allows more of the already existing FPs to be included, thus promoting their recombination, i.e., impedes their accumulation when concentrations are high. These subcascades lower the recombination of FPs at low FPs concentrations but inhibit their accumulation at high FPs concentrations. The presence of CSRO is also beneficial in inhibiting the growth of point defect clusters, which further improves the resistance of HfNbTaTiZr to dislocation generation. Furthermore, the presence of CSRO facilitates the irradiation-induced phase transition. But it is found that HfNbTaTiZr shows suppression of HCP cluster growth. And the tendency to break down large HCP clusters into smaller ones is demonstrated in the CSRO model. From our calculations we also find that the irradiation-induced HCP atoms have a higher potential energy relative to the matrix. The potential energy difference between those energetic HCP atoms and the matrix can lead to generating a great number of insurmountable barriers pervading the matrix and largely suppressing the long-term mobility of FPs, thus limiting their aggregation and growth into clusters.

Improved approaches for small punch test to estimate the yield and ultimate tensile strength of metallic materials

The small punch test (SPT) has been used extensively in nuclear industries to estimate mechanical properties changes of metals due to irradiation. This paper aims to propose more accurate approaches for SPT to estimate the yield and ultimate tensile strength of metals. The relationship between the SPT force and the yield and ultimate tensile strength of materials was studied by using finite element simulation data, and then three approaches were developed to determine tensile properties of materials by means of SPT. Force method and Slop method were developed to derive the ultimate tensile strength of metals from SPT curves, and Area method was proposed to determine the yield and ultimate tensile strength of metals. The accuracy of these approaches were verified by two candidate structural material for fusion reactors (CLF-1 steel and 9Cr-ODS steel) and other 10 structural materials. Area method gives more accurate evaluation on the yield strength of metals than the correlation method given in European standard on SPT. In addition, the developed approaches have advantages in estimating the tensile properties of low ductility metals.

A First-Principles Study of the Structural and Thermo-Mechanical Properties of Tungsten-Based Plasma-Facing Materials

Tungsten (W) and tungsten alloys are being considered as leading candidates for structural and functional materials in future fusion energy devices. The most attractive properties of tungsten for the design of magnetic and inertial fusion energy reactors are its high melting point, high thermal conductivity, low sputtering yield, and low long-term disposal radioactive footprint. Despite these relevant features, there is a lack of understanding of how the structural and mechanical properties of W-based alloys are affected by the temperature in fusion power plants. In this work, we present a study on the thermo-mechanical properties of five W-based plasma-facing materials. First-principles density functional theory (DFT) calculations are combined with the quasi-harmonic approximation (QHA) theory to investigate the electronic, structural, mechanical, and thermal properties of these W-based alloys as a function of temperature. The coefficient of thermal expansion, temperature-dependent elastic constants, and several elastic parameters, including bulk and Young’s modulus, are calculated. Our work advances the understanding of the structural and thermo-mechanical behavior of W-based materials, thus providing insights into the design and selection of candidate plasma-facing materials in fusion energy devices.

Characterization of irradiation-induced dislocation loops in Vanadium at 25-500 °C

Vanadium-based alloys have emerged as promising candidates for structural materials in fusion applications. However, as its base metal, the response of V to irradiation has received limited attention in prior studies. To gain a fundamental understanding of the irradiation damage in V, its microstructure evolution under 6 MeV Ti ion irradiation at 25–500 °C is investigated in the present study, with the focus on the detailed and comprehensive characterization of the behavior of the irradiation-introduced dislocation loops. Under room temperature irradiation, the “black dot” dislocation loops agglomerate linearly into rafts, during which their Burgers vectors are well aligned. With the temperature increases to 300 °C, the size of rafts increases and the density decreases, while the size of small loops maintains similar to the room temperature irradiation condition. As the irradiation temperature reaches 500 °C, the defects become highly mobile, resulting in the formation of extended dislocation loops or lines with hundreds of nanometers in size, with the rafts vanishing. All the observable loops under this irradiation temperature range exhibit the Burgers vectors of a/2 < 111>. All the loops observed in the displacement region are identified to be interstitial-type, while a small portion of loops observed in the diffusion region under elevated temperatures are vacancy-type.

Influence of dimensional deviation and thickness non-uniformity on the small-size specimen tensile results of CLF-1 steel

Tensile testing of small-size specimens has been widely used to evaluate the influence of neutron irradiation on the mechanical properties for structural materials of fusion reactors. In recent years, researchers have been trying to standardize small-size specimen tensile test method. In the present paper, we have investigated the influence of dimensional deviation and thickness non-uniformity on the tensile results of small-size specimen by both experiments and finite element simulations, based on one of the candidate structural materials of fusion reactor, i.e. CLF-1 steel. It was found that under the conditions of small grain size, small metallurgical defect size and suitable specimen preparation, the tensile results of SS-J3 specimen could be close to the results of large-size specimen. Slightly higher yield and ultimate strengths for small specimens may be caused by the surface treatment of the wheel grinding. The repeatability of ultimate strength was better than that of yield strength, and the repeatability of elongation was worse than that of strength. For tensile tests on eight SS-J3 specimens, the maximum difference in total elongation was about 3 %. With ±0.1 mm deviation of thickness and parallel width, the maximum variation in total elongation was about 1.8 %. The ductility properties were sensitive to specimen's thickness non-uniformity, the measured uniform and total elongations would obviously decrease even with 0.01 mm thickness non-uniformity.

A single-phase Nb25Ti35V5Zr35 refractory high-entropy alloy with excellent strength-ductility synergy

Refractory high entropy alloys (RHEAs) are promising candidates for high-temperature structural materials due to their excellent high-temperature performance. However, the room-temperature brittleness of most alloys results in poor plastic formability, thereby limiting their engineering applications. In this study, Nb25Ti35V5Zr35 alloy with cold-rollability was developed, and its phase stability, tensile mechanical properties, elastic properties, strengthening mechanism, and deformation mechanism were investigated by combining experiments, CALPHAD, DFT and molecular statics (MS) methods. The results demonstrate that Nb25Ti35V5Zr35 alloy possesses good phase stability, with both the as-cast and annealed states maintaining a single-phase BCC structure without undergoing phase transformation. The tensile yield strength of the alloy reaches its peak at 1152 MPa in the cold-rolled condition, while its annealed state exhibits an optimal balance of strength and plasticity, with a yield strength of 836 MPa and an elongation of 22.45 %, respectively. Solid solution strengthening is the primary source of its high strength, with V and Zr playing key roles in this strengthening mechanism. Dislocation slip plays a dominant role in plastic deformation. Additionally, compared to traditional BCC alloys, the significance of edge dislocations in the deformation process is notably enhanced, with the {112} plane serving as the preferred slip plane for dislocations. DFT results indicate that the alloy exhibits significant anisotropy in both Young's modulus and shear modulus. This work contributes to understanding the material properties of the Nb25Ti35V5Zr35 alloy and provides a reference for the design of new ductile RHEAs.

Oxidation Behavior of a Novel Nickel-Based Alloy in Air and Steam at 1273 K for the Oxygen–Hydrogen Combustion Chamber

AbstractAs part of advancing oxygen–hydrogen combustion power generation technology, a study was carried out to evaluate the oxidation behavior of a novel developed Ni–Cr–W alloy as the structural material candidate. Tungsten is utilized in the alloy as a solid solution-strengthened element and as an α2-W precipitate former. The examination involved exposing the developed alloy and commercial alloys, Hastelloy X and Nimonic 263, to air and steam environments at 1273 K. The results show a different oxidation behavior of the developed alloy. Considering the air oxidation kinetics, the performance of the developed alloy was on par with that of Hastelloy X and superior to Nimonic 263. A single outer chromia scale was established with an intergranular oxide. Whereas steam exposure resulted in the formation of outer and inner chromia scales with a deeper intergranular oxide penetration. Thicker chromia formation with a lower mass gain indicates the evaporation of chromia under a steam atmosphere.

Influence of He injection amount on the microstructure of ion irradiated Cr0.8FeMnNi medium entropy alloy compared with 316L stainless steel

In recent years, medium and/or high-entropy alloys (M/HEAs) have attracted attention due to their excellent mechanical properties, positioning them as potential candidates for structural materials in advanced nuclear reactors. This study investigated the effect of He injection amount on the microstructural evolution in Cr0.8FeMnNi MEA and 316L stainless steel. Helium atoms were injected, followed by Fe2+ irradiation. Transmission electron microscopy was used to characterize the irradiation-induced defects. The primary irradiation-induced defects observed across all He injection amounts were black dots; no voids were seen under any irradiation conditions. In He-injected Cr0.8FeMnNi, the number density of black dots was significantly lower than in non-He-injected Cr0.8FeMnNi, while the average size of black dots increased with higher He injection rates. As the He injection amount increased, the binding energy of the self-interstitial atom (SIA) to the He-V complex decreased. This reduction facilitated the nucleation and growth of numerous SIA clusters within the matrix, consequently increasing the number density of black dots. The change in the number density of black dots in Cr0.8FeMnNi was much smaller than that observed in 316L. This could be ascribed to the enhanced self-healing process caused by the diversity of point defects and higher lattice distortion in Cr0.8FeMnNi.

Microstructure evolution and impact toughness degradation of a 3.6%Si austenitic stainless steel during high-temperature exposure

Si-modified Fe–Cr–Ni austenitic stainless steel is a promising candidate structural material in lead-cooled fast reactor systems and high-temperature microstructural stability is the challenge for developing this kind of materials. In the present research, the microstructure and corresponding impact toughness evolutions of a 3.6 wt%Si austenitic stainless steel (3.6Si-ASS) aged between 510 and 650 °C up to 5000 h were investigated. It was found that the microstructure showed nearly no change within 2000 h at 510 °C. Subsequently, the G-phase (Cr3Ni2Si) began to form at grain boundaries after 3000 h. This made the matrix near grain boundaries Ni-depleted and promoted the formation of ferrite. Finally, a mixed microstructure consisting of G-phase and ferrite was formed after 5000 h. The appearance of ferrite caused a slight decrease in impact toughness. This evolution was further accelerated during aging at 550 °C, and the impact toughness showed a significant decrease when ferrite appeared. Interestingly, the G-phase and ferrite were suppressed, replaced by intergranular M6C ((Cr, Mo)3Ni2SiC) carbide and intragranular (Cr, Si, Mo)-rich χ-phase during aging at 600 and 650 °C. Herein, M6C carbides preferred to form at grain boundaries, making intergranular cracking easier and reducing the toughness. When aged at 650 °C, the higher diffusion rates of Cr, Ni, Si and Mo increase grain boundaries coverage of M6C and also the number density, diameter, and area fraction of the χ-phase.

Simulations of radiation damage accumulation in Fe-9Cr under pulsed irradiation conditions representative of inertial fusion energy

Structural materials for laser-based inertial fusion energy (IFE) reactor concepts are expected to operate under pulsed irradiation conditions, with cycles consisting of microsecond-long neutron bursts followed by inter-pulse periods of up to one second in duration. During each laser shot, irradiation damage is introduced at dose rates that are up to six orders of magnitude higher than those in their magnetic fusion energy (MFE) counterparts. Under certain conditions, the inter-pulse periods may last an amount of time sufficient to anneal much of the damage introduced during each shot. This phenomenon is highly temperature dependent, with pulsed heating directly linked to the pulsed damage, large surface temperature spikes may also occur. As such, this intermittent mode of operation has the potential to lead to fundamental differences in how irradiation damage accumulates in structural reactor materials. However, damage to structural materials under IFE conditions has received comparatively much less attention than in MFE, as pulsed conditions add yet an extra dimension to the already extremely challenging problem of microstructural evolution under fusion neutron irradiation in structural materials. In this work we use the stochastic cluster dynamics (SCD) method to simulate the evolution with time of defect cluster concentrations under IFE conditions. We consider the Laser Inertial Fusion Energy (LIFE) reactor concept as the representative IFE design for our study, for which detailed spectral information is available, including gas transmutant production. We simulate several pulse frequencies and three different temperatures, and compare the results with continuous irradiation cases under identical average dose rates. The simulations are run in Fe-9Cr system as a model alloy for reduced-activation ferritic/martensitic (RAFM) steels, which are the leading structural material candidates for first-wall structures in MFE and IFE devices. We find that, in practically all scenarios, pulsed irradiation restricts the formation of helium-vacancy clusters relative to the levels seen under equivalent steady irradiation conditions. As well, although self-interstitial atom clusters do accumulate under pulsed operation, their number densities remain up to an order of magnitude lower than in continuous irradiation conditions. Based on the SCD results, we provide a temperature-pulse rate map to identify regions where pulsed irradiation may lead to larger defect accumulation than under continuous irradiation.

Creep behavior of a novel ODS ferrite steel reinforced with ultra-fine Y2(Zr0.6, Ti0.4)2O7 particles

Oxide dispersion strengthened (ODS) ferrite steel is a competitive candidate structural material for the Generation IV fission reactors. In this work, a novel ODS ferrite steel reinforced with ultra-fine Y2(Zr0.6, Ti0.4)2O7 particles was prepared by powder metallurgy method, which exhibit superior mechanical properties at elevated temperatures. Creep tests of the ODS ferrite steel at 800 °C was conducted to reveal its high-temperature creep failure mechanism. This ODS steel has an average grain size of 1.31 μm, and the mean size and number density of the ultra-fine oxide particles are 5.48 nm and 4.85 × 1023 m−3, respectively. When the applied creep stress is 100 MPa, the creep rupture time of the ODS steel is 2448.5 h, and the minimum creep rate is only 1.05 × 10−7 s−1. Under the similar creep stress at 800 °C, the creep rate of this ODS steel is at least one order of magnitude lower than that of other typical ODS steels reported in the literatures, showing excellent creep properties. By fitting the data of minimum creep rate, applied stress and effective stress normalized by shear modulus, the creep threshold stress of the ODS steel was determined as 56.2 MPa, and the true stress exponent was calculated as 4.5 ± 1.0, which indicates that the creep mechanism is dominated by dislocation. Orowan bypass mechanism and dislocation climb over the particle mechanism play a major role in this dislocation mediated creep behavior. Due to the strong hindering effect of dislocation motion caused by the ultra-fine and dispersed Y2(Zr0.6, Ti0.4)2O7 particles, the novel ODS ferrite steel thereby achieved excellent creep properties.

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

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

Ultra-Fast-Healing Glassy Hyperbranched Plastics Capable of Restoring 26.4 MPa Tensile Strength within One Minute at Room Temperature.

The growing concern regarding widespread plastic pollution has propelled the development of sustainable self-healing plastics. Although considerable efforts have been dedicated to fabricating self-healing plastics, achieving rapid healing at room temperature is extremely challenging. Herein, we have developed an ultra-fast-healing glassy polyurethane (UGPU) by designing a hyperbranched molecular structure with a high density of multiple hydrogen bonds (H-bonds) on compliant acyclic heterochains and introducing trace water to form water bridge across the fractured surfaces. The compliant acyclic heterochains allow the dense multiple hydrogen bonds to form a frozen network, enabling tensile strength of up to 70 MPa and storage modulus of 2.5 GPa. The hyperbranched structure can drive the reorganization of the H-bonding network through the high mobility of the branched chains and terminals, thereby leading to self-healing ability at room temperature. Intriguingly, the presence of trace water vapor facilitates the formation of activated layers and the rearrangement of networks across the fractured UGPU sections, thereby enabling ultra-fast self-healing at room temperature. Consequently, the restored tensile strength after healing for 1 minute achieves a historic-record of 26.4 MPa. Furthermore, the high transparency (>90 %) and ultra-fast healing property of UGPU make it an excellent candidate for advanced optical and structural materials.

Ultra‐Fast‐Healing Glassy Hyperbranched Plastics Capable of Restoring 26.4 MPa Tensile Strength within One Minute at Room Temperature

AbstractThe growing concern regarding widespread plastic pollution has propelled the development of sustainable self‐healing plastics. Although considerable efforts have been dedicated to fabricating self‐healing plastics, achieving rapid healing at room temperature is extremely challenging. Herein, we have developed an ultra‐fast‐healing glassy polyurethane (UGPU) by designing a hyperbranched molecular structure with a high density of multiple hydrogen bonds (H‐bonds) on compliant acyclic heterochains and introducing trace water to form water bridge across the fractured surfaces. The compliant acyclic heterochains allow the dense multiple hydrogen bonds to form a frozen network, enabling tensile strength of up to 70 MPa and storage modulus of 2.5 GPa. The hyperbranched structure can drive the reorganization of the H‐bonding network through the high mobility of the branched chains and terminals, thereby leading to self‐healing ability at room temperature. Intriguingly, the presence of trace water vapor facilitates the formation of activated layers and the rearrangement of networks across the fractured UGPU sections, thereby enabling ultra‐fast self‐healing at room temperature. Consequently, the restored tensile strength after healing for 1 minute achieves a historic‐record of 26.4 MPa. Furthermore, the high transparency (&gt;90 %) and ultra‐fast healing property of UGPU make it an excellent candidate for advanced optical and structural materials.

Unveiling the deformation micro-mechanism for mechanical anisotropy of a CoCrFeNi medium entropy alloy

The equiatomic Cr-Co-Fe-Ni medium-entropy alloy has the face-centered cubic structure. Single crystals of this alloy were tested by in-situ micropillar compression along different loading axes under scanning electron microscope. The transmission electron microscopy characterization and molecular dynamics simulation were incorporated for quantitative analysis of the effects of different crystal orientations on the deformation mechanisms. The <001>-oriented pillar not only exhibited extensive deformation-induced nano twinning, but also has been identified for the first time to undergo the FCCHCP phase transformation at room temperature. The strain localization tendency of <011>-oriented samples was confirmed through uniaxial tests to interpret the significant serration on stress-strain curves. The prominent strain hardening of <111>-oriented pillars was attributed to intense intersection between slip planes as evidenced by the extra density of Lomer-Cottrell locks. Such a high hardening rate has caused subsequent kinking of pillars. Functional division of different regions of kink band was conducted based on Orowan model. In principle, multi-principal element alloys can theoretically be designed and developed to combine a variety of excellent properties, which is an important class of candidate structural materials for advanced engineering systems. These findings provide promising guidance for understanding the mechanical anisotropy and application of these alloys.

Effect of cooling rate on the precipitation characteristics and growth mechanism of Bi3Ni in liquid lead‑bismuth eutectic

Austenitic stainless steel is a promising candidate structural material for direct contact components with lead‑bismuth eutectic (LBE) in lead-cooled fast reactor (LFR). However, the preferential dissolution of Ni from austenitic steels at low oxygen concentrations will lead to the precipitation of Ni-containing phases in the low-temperature parts of reactors, which has a risk of pipe clogging. In this paper, three samples, comprising one extracted from a corrosion tank after prolonged operation and two prepared, respectively, by quenching in liquid nitrogen and cooling in the furnace, were obtained to investigate the effect of cooling rate on the precipitation characteristics and growth mechanism of Bi3Ni that forms by the reaction of dissolved Ni and Bi in LBE. The results showed that rod-like Bi3Ni was found in all three Ni-containing LBE samples. As the cooling rate increased, the width of Bi3Ni decreased. The Bi3Ni in all three samples displayed a pseudo-hexagonal prismatic shape and had a typical faceted growth character; the cooling rate had no effect on the morphology of them. The crystallographic analysis showed that the growth axis of Bi3Ni was in the 〈010〉 direction, and the prism planes were composed of {001} and {101}. In addition, the precipitation of Bi3Ni inevitably increased the content of Pb7Bi3, which would in turn enhance the volumetric expansion of solid LBE samples. For the growth mechanism, Bi3Ni phases were controlled by the screw dislocation mechanism, with the step growth rate of the dislocation core being greater than that of the outer edges. Given the low cooling rate and low Ni content in the LBE coolant of reactors, the precipitation of Bi3Ni with a rod-like morphology seems to be inevitable.

AlxHfTaTi (0≤x≤0.5) refractory medium entropy alloys with excellent room-temperature tensile properties

Refractory high entropy alloys are strong candidates for high-temperature structural materials, because of their high strength, excellent thermal stability and exceptional softening resistance under high-temperature. However, majority of the existing refractory high entropy alloys are seriously limited in widespread application, due to their severe brittleness at room temperature. In this study, we proposed a novel AlxHfTaTi body-centered cubic (BCC) refractory medium entropy alloy with remarkable strength-ductility combination at room temperature. Adjusting the atomic ratio of Al (from 0 to 0.5 at.%), we investigated the effect of Al on the structure-property characteristics of the alloy under as-cast as well as cold-rolled and annealed conditions. Compared with the as-cast samples, the rolled and annealed samples exhibit more excellent tensile properties due to the remarkable reduction of grain size resulted from recrystallization. The Al0.3HfTaTi sample annealed for 5 min exhibits outstanding comprehensive properties at room temperature, with a yield strength higher than 1000 MPa and a fracture elongation of up to 24 ± 1 %, respectively. Quasi in-situ EBSD analysis results reveal that, during tensile deformation, the grain size has a significant effect on the grain rotation and dislocation gliding. Small grains are more likely to initiate multiple slip bands, promoting the synergistic effect of grain rotation and dislocation gliding, thereby improving the ductility of the alloy. Our findings can provide new ideas for the development of refractory high entropy alloys with outstanding room-temperature tensile properties.

Microstructure evolution and oxidation resistance properties of oxide dispersion strengthened FeCrAl alloys at 1100 °C

Oxide dispersion strengthened (ODS) FeCrAl alloys were prepared by mechanical alloying (MA) and spark plasma sintering (SPS) technology. To reveal the high temperature oxidation properties, ODS FeCrAl alloys were exposed to dry air for various times (1h, 4h, 10h, 20h, 45h, 70h, and 100h) at 1100 °C to investigate the microstructure evolution. The results illustrated that a flat and adherent alumina scale was gradually thickened on the alloy substrate with the increase of the oxidation time. In particular, it was observed that large-size reactive element (RE) oxides were inclined to be associated with cavities close to the gas interface and these cavities were progressive evolution to be larger-size pores with longer exposure time, while tiny RE oxides were not dispersed with cavities after 100h oxidation. In addition, α-Al2O3 with a double-scale structure were mainly composed of inner columnar grains and outer equiaxed grains, as well, the growth of α-Al2O3 scales illustrated that the oxidation rate decreased with oxidation reaction proceeding. For one thing, Ti segregation along grain boundaries (GBs) near the gas interface decreased the outward diffusion of the Al ions to weaken the oxidation rate. Furthermore, the gradual coarsening of columnar grains near the alloy substrate implied a longer diffusion path of O atoms, thereby accounting for slower oxidation process. To sum up, ODS FeCrAl alloys with thinner oxide thickness demonstrated the superior oxidation resistance than other materials, which would make it one of the promising candidate structural materials applicated in accident-tolerant fuel (ATF) cladding materials.

Chromium agglomeration induced by Fe+ ion irradiation of Fe-10at%Cr

Fe-Cr alloys serve as model alloys for the investigation of radiation induced effects in ferritic-martensitic steels which are candidate structural materials for future fusion reactors. In this work the effect of Cr segregation and/or agglomeration in 490 keV Fe+ ion irradiated Fe-10at%Cr alloys in the form of thin films is investigated. The irradiations took place at 300 °C at doses ranging from 0.5 to 20 displacements per atom (dpa). Polarized Neutron Reflectivity (PNR) measurements were used for the determination of the solute Cr concentration in the Fe-Cr matrix. Cr depletion from the Fe-Cr matrix up to 2.4 at% was found. This is related to solute Cr decrement as the accumulated dose increases. After the damage of 4 dpa, solute Cr reaches the asymptotic value of 8.4 at%, close to that of the thermodynamic equilibrium in Fe-Cr. Atom Probe Tomography (APT) measurements showed that after irradiation Cr accumulates into clusters the majority of which is co-located with oxygen.

Corrosion Behavior of Candidate Structural Materials for Molten Salt Reactors in Flowing NaCl-MgCl2

Abstracts. A high-temperature molten salt natural convection loop was designed and manufactured for chloride-based salts. NaCl-MgCl2 salt was prepared and injected into the loop. Corrosion coupons composed of SS304, SS316L, and high-Ni alloys were also prepared as candidate structural materials for molten chloride salt reactors. After a corrosion experiment was conducted for 500 h, the salt was drained from the loop to the drain tank; inductively coupled plasma optical emission spectroscopy was used to investigate metal dissolution from the materials into the salt. The corroded materials were analyzed using scanning electron microscopy and energy-dispersive X-ray spectroscopy to calculate the corrosion rate. Materials exposed to NaCl-MgCl2 molten salt showed varying levels of corrosion resistance. The high-Ni alloy demonstrated the highest resistance, followed by SS316L and SS304. Furthermore, corrosion products were observed to migrate along the molten salt through natural convection, eventually depositing onto the surface of the high-Ni alloy in the cold leg of the loop.