Sinks For Defects Research Articles

The effects of loop size on the unfaulting of Frank loops in heavy ion irradiation

When stainless steels are exposed to high irradiation dose, a fraction of Frank loops will unfault and transform into perfect loops. After unfaulting, perfect loops can glide and impinge into other loops to form dislocation networks, and dislocation networks will act as sinks for point defects and solute elements. Besides, unfaulting will influence the extent of irradiation hardening as well. Based on previous theories, the size of Frank loops was considered to be a key factor in unfaulting. To better understand the unfaulting process, 316L stainless steel model alloy was irradiated to 5dpa at 350 °C, 400 °C and 450 °C by 3 MeV Ni ions. Method was developed to quantitatively compare the size and density between Frank loops and perfect loops through relrod and weak-beam dark field techniques. Results showed that unfaulting barely occurred at 350 °C but was distinct at 400 °C and 450 °C. The majority of perfect loops were in the size range of 8–16 nm at 400 °C and were in the size range of 12–20 nm at 450 °C. The existence of a critical loop size for unfaulting was not supported by our results. Moreover, by our results, the energy of Frank loops does not play a decisive role in unfaulting, and the simple approximation of geometric constraint in the rate theory to calculate unfaulting diameter needs to be revisited. Irradiation hardening was also analyzed by nano-indention. It was found that loop unfaulting played an important role in the hardening decrease from 350 °C to 450 °C.

Modern nanostructured ferritic alloys: A compelling and viable choice for sodium fast reactor fuel cladding applications

Sodium fast reactor cores present a truly challenging environment for the fuel cladding and core structural materials that limit achievable fuel burnup in these systems. Modern nanostructured ferritic alloys have the potential to deliver the desired high performance characteristics where historic austenitic and ferritic/martensitic alloys fall short. In this paper, a new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), is developed and demonstrated in cladding geometry with a length > 1 m. The alloy composition and microstructure are tailored to deliver dramatically improved strength and creep resistance. At 600 °C the ultimate tensile strength is ∼600 MPa and the stress to induce a steady state strain rate of 10−6 s−1 is 500 MPa vs. 200 MPa for traditional ferritic/martensitic alloys. A very high defect sink density was engineered in these alloys that is expected to further enhance the traditional good swelling and irradiation creep resistance of ferritic/martensitic steels. These significant improvements along with the demonstrated viability for seamless cladding production indicate that this alloy is a compelling and viable choice for sodium fast reactor fuel cladding applications.

Defect accumulation and evolution during prolonged irradiation of Fe and FeCr alloys

The understanding of materials’ behaviour during continuous irradiation is of great interest for utilizing materials in environments where harsh radiation is present, like nuclear power plants. Most power plants, both current and future ones, are based, at least partially, on Fe or FeCr alloys. In this study, we investigate the response of BCC Fe and several FeCr alloys to massively overlapping cascades. The effect of the added chromium on the defect accumulation and defect evolution was studied. Both a bulk setup, for observing the evolution deep inside the material far from grain boundaries and surfaces, and a setup with a nearby open surface, to investigate the effect of a permanent defect sink, were studied. We found that the primary defect production is similar in all materials, and also the build-up before serious overlap is comparable. When cascade overlap starts, we found that different sized clusters are formed in the different materials, depending on the setup. The defect cluster evolution was followed and could be related to the dislocation reactions in the materials. We found that the irradiation mixing was specific to the different chromium concentrations, the low chromium-containing alloy showed ordering, whereas the highest chromium-containing sample showed segregation.

Modelling the influence of strain fields around precipitates on defect equilibria and kinetics under irradiation in ODS steels: A multi scale approach

Resistance to radiation damage is a crucial property for structural materials in fission or fusion reactors. The particle-matrix interface in oxide dispersion strengthened (ODS) steels is considered a sink for point defects created during irradiation. In this work we address the question, if elastic strain fields around oxide particles cause a long-range interaction between the precipitates and point defects. We use kinetic Monte Carlo simulations to simulate the diffusion of point defects under the influence of elastic strain fields caused by Y2O3 and Y2Ti2O7 precipitates. We show that there is essentially no interaction between vacancies and the strain fields, while the sink strength of precipitates for interstitials increases with misfit strain between precipitate and matrix. Consequently, the interstitial concentration decreases with misfit strain, while due to reduced interstitial-vacancy recombination the vacancy concentration increases with misfit strain. The total change of point defect concentration with misfit strain is, however, rather limited.

Reduction of defect generation and development of sinks at nanocluster boundary in oxide dispersion-strengthened steel

The radiation resistance mechanisms of nanoclusters (NCs) in oxide dispersion-strengthened (ODS) steels have been investigated. Molecular dynamics simulation has been used to investigate defect generation during the primary damage state of a displacement cascade in ODS steels for NCs of various radii and a range of primary knock-on atom (PKA) energies. Y2O3 NCs considerably enhance the radiation resistance of ODS steels by reducing the peak defect generation during the cascade within the Fe matrix. The NC also affects the morphology of the collision cascades, depending on PKA energy. At lower energies, the NC’s outer circumference act as a cessation point forming a dampened shockwave compared to a pure Fe system. At higher energies, the PKA energy is able to transfer through the NC, thus causing two smaller shockwaves in the Fe matrix. Along with the alteration of the cascade morphology, the NC boundary acts as a strong defect sink to absorb defects and defect clusters, leading to significant recombination of interstitials and vacancies away from the NC. The interfacial energy of the NCs with the Fe matrix increases with increasing diameter of the oxide NCs. The evolution of the NC is tracked through the primary damage state of a cascade, and the effects of ballistic dissolution play a key role in this evolution, most evident in the 2 nm NC.

Different Radiation Tolerances of Ultrafine-Grained Zirconia-Magnesia Composite Ceramics with Different Grain Sizes.

Developing high-radiation-tolerant inert matrix fuel (IMF) with a long lifetime is important for advanced fission nuclear systems. In this work, we combined zirconia (ZrO2) with magnesia (MgO) to form ultrafine-grained ZrO2–MgO composite ceramics. On the one hand, the formation of phase interfaces can stabilize the structure of ZrO2 as well as inhibiting excessive coarsening of grains. On the other hand, the grain refinement of the composite ceramics can increase the defect sinks. Two kinds of composite ceramics with different grain sizes were prepared by spark plasma sintering (SPS), and their radiation damage behaviors were evaluated by helium (He) and xenon (Xe) ion irradiation. It was found that these dual-phase composite ceramics had better radiation tolerance than the pure yttria-stabilized ZrO2 (YSZ) and MgO. Regarding He+ ion irradiation with low displacement damage, the ZrO2–MgO composite ceramic with smaller grain size had a better ability to manage He bubbles than the composite ceramic with larger grain size. However, the ZrO2–MgO composite ceramic with a larger grain size could withstand higher displacement damage in the phase transformation under heavy ion irradiation. Therefore, the balance in managing He bubbles and phase stability should be considered in choosing suitable grain sizes.

Interface Effects on He Ion Irradiation in Nanostructured Materials

In advanced fission and fusion reactors, structural materials suffer from high dose irradiation by energetic particles and are subject to severe microstructure damage. He atoms, as a byproduct of the (n, α) transmutation reaction, could accumulate to form deleterious cavities, which accelerate radiation-induced embrittlement, swelling and surface deterioration, ultimately degrade the service lifetime of reactor materials. Extensive studies have been performed to explore the strategies that can mitigate He ion irradiation damage. Recently, nanostructured materials have received broad attention because they contain abundant interfaces that are efficient sinks for radiation-induced defects. In this review, we summarize and analyze the current understandings on interface effects on He ion irradiation in nanostructured materials. Some key challenges and research directions are highlighted for studying the interface effects on radiation damage in nanostructured materials.

Key factors in radiation tolerance of BCC metals under steady state

Different materials would exhibit discrepant behaviors of radiation tolerance under energetic particle irradiation, due to the effect of intrinsic features (like defect migration energy and sink strength of grain boundaries) and extrinsic conditions (like irradiation rate and temperature) on defect accumulation. We wonder what the general rules of radiation tolerance are in body-centered cube (BCC) transition metals. Based on the steady state chemical rate theory, the influence of these intrinsic features and extrinsic conditions on vacancy accumulation is thus investigated in typical BCC metals including V, Cr, Fe, Nb, Mo, Ta and W. It shows that V, Cr, Fe and Nb exhibit obvious advantage in radiation resistance than that of Ta, Mo and W under typical service conditions, due to their higher vacancy diffusivities. In addition, the smaller grain size and lower irradiation rate, the higher anti-irradiation ability of BCC metals. Therefore, we propose that, under practical irradiation rates and temperatures, BCC metals with low vacancy migration energy and small grain size should be recommended in the selection of new nuclear materials, at least in the view of steady state.

Bi-metal interface-mediated defects distribution in neon ion bombarded Cu/Ag nanocomposites

Bi-metal interfaces act as sinks for radiation defects and hence play an important role in the design of radiation-resistant materials. Here, we report that under neon ion bombardment, the hetero-twin Cu/Ag interface serves as a vacancy pump, transferring vacancies from Cu to Ag, a mechanism first observed under helium radiation (Acta Materialia 160 (2018) 211–223). By comparing helium and neon ion irradiated samples, we reveal that helium likes to aggregate with vacancies and slow down the recombination rate of radiation-induced interstitials and vacancies, which leads to significant swelling, radiation-enhanced diffusion and a stronger vacancy pump effect.

Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: effect of dose

A ferritic/martensitic steel, HT9, was irradiated in the BOR-60 reactor to ∼17.1 and ∼35.1 displacements per atom (dpa) at 650 ± 23 K (377 ± 23 °C). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the role of irradiation dose on the microstructure evolution. Radiation-induced Mn/Ni/Si-rich (G-phase) and radiation-enhanced Cr-rich (αʹ) precipitates were observed within the martensitic laths at all doses. In addition, the G-phase was also observed to precipitate heterogeneously at various defect sinks. The number density for these second-phase precipitates decreases while the size increases with increasing dose, resulting in an increase of the volume fraction. Both a <100> and a/2 <111> type loops were observed with the a <100> type being the predominant type at both doses. The proportion of a <100> loops is consistent with that previously observed in HT9 ion irradiated to similar doses at ∼693–743 K (∼420-470 °C). Only small cavities (diameter < 2 nm) were observed at ∼17.1 dpa whereas both small and large cavities were observed at ∼35.1 dpa, resulting in a bi-modal cavity size distribution at this dose. Alloying elements, Ni and Si, were observed to segregate to the cavity surface, forming Ni/Si-rich shells around the cavities. The swelling at ∼17.1 dpa is evaluated at ∼0.02% while the swelling at ∼35.1 dpa is found to be ∼0.07% with variations from grain to grain. attributed to the spatial variation of the density of large cavities (in different grains). The swelling data obtained in this study was compared with the neutron data of F/M steels available in the literature.

The Effect of Internal Free Surfaces on Void Swelling of Irradiated Pure Iron Containing Subsurface Trenches

We studied the effects of internal free surfaces on the evolution of ion-induced void swelling in pure iron. The study was initially driven by the motivation to introduce a planar free-surface defect sink at depths that would remove the injected interstitial effect from ion irradiation, possibly enhancing swelling. Using the focused ion beam technique, deep trenches were created on a cross section of pure iron at various depths, so as to create bridges of thickness ranging from 0.88 μm to 1.70 μm. Samples were then irradiated with 3.5 MeV Fe2+ ions at 475 °C to a fluence corresponding to a peak displacement per atom dose of 150 dpa. The projected range of 3.5 MeV Fe2+ ions is about 1.2 μm so the chosen bridge thicknesses involved fractions of the ion range, thicknesses comparable to the mean ion range (peak of injected interstitial distribution), and thicknesses beyond the full range. It was found that introduction of such surfaces did not enhance swelling but actually decreased it, primarily because there were now two denuded zones with a combined stronger influence than that of the injected interstitial. The study suggests that such strong surface effects must be considered for ion irradiation studies of thin films or bridge-like structures.

Injection of oxygen vacancies in the bulk lattice of layered cathodes.

Surfaces, interfaces and grain boundaries are classically known to be sinks of defects generated within the bulk lattice. Here, we report an inverse case by which the defects generated at the particle surface are continuously pumped into the bulk lattice. We show that, during operation of a rechargeable battery, oxygen vacancies produced at the surfaces of lithium-rich layered cathode particles migrate towards the inside lattice. This process is associated with a high cutoff voltage at which an anionic redox process is activated. First-principle calculations reveal that triggering of this redox process leads to a sharp decrease of both the formation energy of oxygen vacancies and the migration barrier of oxidized oxide ions, therefore enabling the migration of oxygen vacancies into the bulk lattice of the cathode. This work unveils a coupled redox dynamic that needs to be taken into account when designing high-capacity layered cathode materials for high-voltage lithium-ion batteries.

Point defect sink strength of low-angle tilt grain boundaries: A phase field dislocation climb model

Evaluating the point defect sink strength of grain boundaries is crucial for understanding the metal behavior of plasticity and damage under irradiation. In this paper, the point defect sink strength of low-angle symmetrical tilt grain boundaries is investigated by the phase field dislocation climb model under irradiation. The results indicate that sink strength of grain boundary is not only determined by the long-range point defect diffusion but also the short-range point defect absorption by the dynamic climbing of grain boundary dislocations. All of the study findings prove that the irradiation induced creep deformation of grain boundaries is essential for evaluating the radiation tolerance of materials.

Radiation Tolerance in Nano-Structured Crystalline Fe(Cr)/Amorphous SiOC Composite

The management of irradiation defects is one of key challenges for structural materials in current and future reactor systems. To develop radiation tolerant alloys for service in extreme irradiation environments, the Fe self-ion radiation response of nanocomposites composed of amorphous silicon oxycarbide (SiOC) and crystalline Fe(Cr) were examined at 10, 20, and 50 displacements per atom damage levels. Grain growth in width direction was observed to increase with increasing irradiation dose in both Fe(Cr) films and Fe(Cr) layers in the nanocomposite after irradiation at room temperature. However, compared to the Fe(Cr) film, the Fe(Cr) layers in the nanocomposite exhibited ~50% less grain growth at the same damage levels, suggesting that interfaces in the nanocomposite were defect sinks. Moreover, the addition of Cr to α-Fe was shown to suppress its grain growth under irradiation for both the composite and non-composite case, consistent with earlier molecular dynamic (MD) modeling studies.

Interplay Between Grain Boundaries and Radiation Damage

The need for enhanced radiation-tolerant materials for advanced nuclear energy designs has resulted in a growing number of investigations that have explored the effect of grain boundaries under irradiation. The key motivation for examining the role of grain boundaries in radiation environments is the ability to tailor grain boundary networks through either the introduction of specific grain boundaries or an increase in the grain boundary density. While traditionally thought to be efficient sinks for radiation-induced point defects, many recent experimental studies in model and pure systems have shown significant heterogeneity in grain boundary-defect interactions and associated sink efficiency as a function of grain boundary character. Furthermore, grain boundaries can migrate under irradiation, which creates an additional level of complexity. This article will provide a prospective on the experimental observations associated with defect evolution near grain boundaries including variation in sink efficiency and grain boundary mobility in radiation environments.

An in situ study on Kr ion–irradiated crystalline Cu/amorphous-CuNb nanolaminates

Abstract

Modeling the Influence of Strain Fields Around Precipitates on Defect Equilibria and Kinetics Under Irradiation in ODS Steels: A Multi Scale Approach

Resistance to radiation damage is a crucial property for structural materials in fission or fusion reactors. The particle-matrix interface in oxide dispersion strengthened (ODS) steels is considered a sink for point defects created during irradiation. In this work we address thequestion, if elastic strain fields around oxide particles cause a long-range interaction between the precipitates and point defects. We use kinetic Monte Carlo simulations to simulate the diffusion of point defects under the influence of elastic strain fields caused by Y2O3 and Y2Ti2O7 precipitates. We show that there is essentially no interaction between vacancies and the strain fields, while the sink strength of precipitates for interstitials increases with misfit strain between precipitate and matrix. Consequently, the interstitial concentration decreases with misfit strain, while due to reduced interstitial-vacancy recombination the vacancy concentration increases with misfit strain. The total change of point defect concentration with misfit strain is, however, rather limited.

Chemically-biased diffusion and segregation impede void growth in irradiated Ni-Fe alloys

Recent irradiations of Ni-Fe concentrated solid solution alloys have demonstrated significant improvement of radiation performance. This improvement is attributed to redistribution of the alloying elements near sinks of point defects (voids, dislocations) due to chemically-biased atomic diffusion, where vacancies have preference to migrate via Fe atoms and interstitials via Ni atoms. In Ni-Fe, all sinks are enriched by Ni atoms, which strongly affects further interactions of radiation-produced mobile defects with voids and dislocations, hence void growth and dislocation climb. Ni-decorated sinks interact stronger with interstitial atoms than vacancies, which enhances dislocation loops growth. At the same time, Ni segregation creates Fe-enriched “channels” for vacancy migration out of the damage region to agglomerate in the outer regions, inaccessible to interstitial atoms. Strong effect of chemically-biased diffusion is supported by transmission electron microscope characterization and calls for special attention in designing alloys with desired properties through tuning defect mobilities.

Molecular dynamics investigation of grain boundaries and surfaces in U3Si2

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy benefits from higher thermal conductivity and higher fissile density compared to uranium dioxide (UO2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multi-scale modeling efforts are underway to address this gap in knowledge. Interfaces play a critical role in the microstructural evolution of nuclear fuel under irradiation, acting both as sinks for point defects and as preferential nucleation sites for fission gas bubbles. In this study, a semi-empirical modified Embedded-Atom Method (MEAM) potential is utilized to investigate grain boundaries and free surfaces in U3Si2. The interfacial energy as a function of temperature is investigated for ten symmetric tilt grain boundaries, eight unique free surfaces and voids of radius up to 35 Å. The point defect segregation energy for both U and Si interstitials and vacancies is also determined for two grain boundary orientations. Finally, the entropy change and free energy change for grain boundaries is calculated as a function of temperature. This is the first study into grain boundary properties of U-Si nuclear fuel.

An Approach to Reachability Determination for Static Analysis Defects with the Help of Dynamic Symbolic Execution

Program analysis methods for error detection are conventionally divided into two groups: static analysis methods and dynamic analysis methods. In this paper, we present a combined approach that allows one to determine reachability for defects found by static program analysis techniques through applying dynamic symbolic execution to a program. This approach is an extension of our previous approach to determining the reachability of specific program instructions by using dynamic symbolic execution. The approach is sequentially applied to several points in the program: a defect source point, a defect sink point, and additional intermediate conditional jumps related to a defect under analysis. Our approach can be briefly described as follows. First, static analysis of the program executable code is carried out to gather information about execution paths that guide dynamic symbolic execution to the source point of a defect. Then, dynamic symbolic execution is performed to generate an input dataset for reaching the defect source point and the defect sink point through intermediate conditional jumps. Dynamic symbolic execution is guided by the heuristic of the minimum distance from the previous path to the next defect trace point when selecting execution paths. The distance metric is computed using an extended call graph of the program, which combines its call graph and portions of its control flow graph that include all paths leading to the defect sink point. We evaluate our approach by using several open-source command line programs from Debian Linux. The evaluation confirms that the proposed approach can be used for classification of defects found by static program analysis. However, we found some limitations that prevent deploying this approach to industrial program analyzers. Mitigating these limitations is one of the possible directions for future research.