Radiation-induced Grain Research Articles

The effects of radiation-induced grain subdivision and dislocations on the fracture properties of uranium dioxide

Microcantilever bending tests were applied on uranium dioxide samples irradiated by 84 MeV Xe ions. Depending on the initial grain sizes, different levels of grain subdivision and dislocation development were observed near the irradiated surface. These radiation damage features were shown to degrade the fracture properties of both samples. The fundamental aspects of the radiation-microstructure-property relationships were discussed in this paper, taking a holistic consideration of the microcantilever bending data, fractography and microstructures. The separate effects of the irradiation-induced grain subdivision and dislocations on the fracture initiation and propagation of uranium dioxide were discussed. It was observed that radiation-induced dislocations and grain subdivision, without the presence of Xe bubbles co-located with the defects, cause reductions in the Young's modulus, fracture strain, and fracture stress, and a transition to intergranular fracture.

Formation of Point Defect Clusters in Metals with Grain Boundaries under Irradiation

Molecular dynamics simulations have been performed to investigate the defect structure evolution at different development stages of atomic displacement cascades with energies up to 50 keV in iron crystallites in the temperature range from 300 to 900 K. The number of surviving radiation defects in iron crystallites increases according to a power law with increasing energy of the primary knocked-on atom. An increase in the crystallite temperature slightly increases the number of surviving defects. It is found that atomic displacement cascades can lead to radiation-induced grain boundary migration due to the melting and crystallization of the radiation-damaged region. The crystallographic orientation of the irradiated free surface strongly affects the radiation damage behavior. Craters with adatom islands are formed on the (111) free surface, and vacancy loops are nucleated in the (110) near-surface region. Point defects aggregate into clusters of various types during the evolution of atomic displacement cascades. It is shown that the number of surviving point defect clusters can significantly decrease under uniaxial elastic compression.

Mechanisms of grain boundary migration and growth in nanocrystalline metals under irradiation

Atomistic simulations of radiation damage uncover how grain boundaries (GBs) migrate and coalesce under irradiation in bicrystalline Cu. Planar GB migration biased by defect cluster-mediated attraction first leads to slow and steady motion. Subsequently, adjoining GBs coalesce into curved surfaces, where curvature-driven migration with a velocity three orders of magnitude higher than that of a planar boundary dominates motion, triggering rapid grain growth. This study reveals the atomistic mechanisms of radiation-induced grain growth, and has practical implications towards engineering radiation-tolerant nanostructures.

Корреляция высокодозового радиационного распухания стали класса 16 Cr – 15 Ni с размером зерна

Корреляция высокодозового радиационного распухания стали класса 16 Cr – 15 Ni с размером зерна

Interaction between Al and atomic layer deposited (ALD) ZrN under high-energy heavy ion irradiation

Uranium-molybdenum (U-Mo) particles dispersed in an aluminum matrix is the most promising candidate fuel to convert high-power research and test reactors in Europe from using high-enriched to using low-enriched fuel. However, chemical interaction between the U-Mo and the Al matrix leads to undesirable fuel behavior. Zirconium nitride (ZrN) is used as a diffusion barrier between the U-Mo fuel particles and the Al matrix. To understand the potential microstructural evolution of ZrN during irradiation, a high-energy heavy ion (84 MeV Xe) irradiation experiment was performed on atomic layer deposited (ALD) nanocrystalline ZrN deposited on an Al plate. A fluence of 1.86×1017 ions/cm2, or 90.3 dpa was reached during this experiment. Both analytic transmission electron microscopy (TEM) and synchrotron microbeam X-ray diffraction (μXRD) techniques were utilized to investigate the kinetics of radiation-induced grain growth of ZrN at various radiation doses based on the Williamson-Hall analyses. The grain growth kinetics can be described by a power law expression, Dn−D0n=Kϕ, with n = 5.1. The Al-ZrN interaction products (Al3Zr and AlN) created by radiation-induced ballistic mixing/radiation-enhanced diffusion and their corresponding formation mechanism were determined from electron diffraction and elemental composition analysis. These experimental results were confirmed by first principle thermodynamic density functional theory (DFT) calculations. The results from this ion irradiation study were also compared to in-pile irradiation data from physical vapor deposited (PVD) ZrN samples for a comprehensive evaluation of the interaction between Al and ZrN and its influence on diffusion barrier performance.

Radiation-induced grain subdivision and bubble formation in U3Si2 at LWR temperature

U3Si2, an advanced fuel form proposed for light water reactors (LWRs), has excellent thermal conductivity and a high fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U3Si2 is available at LWR conditions. This study explores the irradiation behavior of U3Si2 by 300 keV Xe+ ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U3Si2 is stable against radiation-induced amorphization at 350 °C even up to a very high dose of 64 displacements per atom (dpa). Grain subdivision of U3Si2 occurs at a relatively low dose of 0.8 dpa and continues to above 48 dpa, leading to the formation of high-density nanoparticles. Nano-sized Xe gas bubbles prevail at a dose of 24 dpa, and Xe bubble coalescence was identified with the increase of irradiation dose. The volumetric swelling resulting from Xe gas bubble formation and coalescence was estimated with respect to radiation dose, and a 2.2% volumetric swelling was observed for U3Si2 irradiated at 64 dpa. Due to extremely high susceptibility to oxidation, the nano-sized U3Si2 grains upon radiation-induced grain subdivision were oxidized to nanocrystalline UO2 in a high vacuum chamber for TEM observation, eventually leading to the formation of UO2 nanocrystallites stable up to 80 dpa.

Radiation tolerance of nanocrystalline ceramics: insights from Yttria Stabilized Zirconia.

Materials for applications in hostile environments, such as nuclear reactors or radioactive waste immobilization, require extremely high resistance to radiation damage, such as resistance to amorphization or volume swelling. Nanocrystalline materials have been reported to present exceptionally high radiation-tolerance to amorphization. In principle, grain boundaries that are prevalent in nanomaterials could act as sinks for point-defects, enhancing defect recombination. In this paper we present evidence for this mechanism in nanograined Yttria Stabilized Zirconia (YSZ), associated with the observation that the concentration of defects after irradiation using heavy ions (Kr+, 400 keV) is inversely proportional to the grain size. HAADF images suggest the short migration distances in nanograined YSZ allow radiation induced interstitials to reach the grain boundaries on the irradiation time scale, leaving behind only vacancy clusters distributed within the grain. Because of the relatively low temperature of the irradiations and the fact that interstitials diffuse thermally more slowly than vacancies, this result indicates that the interstitials must reach the boundaries directly in the collision cascade, consistent with previous simulation results. Concomitant radiation-induced grain growth was observed which, as a consequence of the non-uniform implantation, caused cracking of the nano-samples induced by local stresses at the irradiated/non-irradiated interfaces.

Competing effects of electronic and nuclear energy loss on microstructural evolution in ionic-covalent materials

Ever increasing energy needs have raised the demands for advanced fuels and cladding materials that withstand the extreme radiation environments with improved accident tolerance over a long period of time. Ceria (CeO2) is a well known ionic conductor that is isostructural with urania and plutonia-based nuclear fuels. In the context of nuclear fuels, immobilization and transmutation of actinides, CeO2 is a model system for radiation effect studies. Covalent silicon carbide (SiC) is a candidate for use as structural material in fusion, cladding material for fission reactors, and an inert matrix for the transmutation of plutonium and other radioactive actinides. Understanding microstructural change of these ionic-covalent materials to irradiation is important for advanced nuclear energy systems.While displacements from nuclear energy loss may be the primary contribution to damage accumulation in a crystalline matrix and a driving force for the grain boundary evolution in nanostructured materials, local non-equilibrium disorder and excitation through electronic energy loss may, however, produce additional damage or anneal pre-existing defects. At intermediate transit energies where electronic and nuclear energy losses are both significant, synergistic, additive or competitive processes may evolve that affect the dynamic response of materials to irradiation. The response of crystalline and nanostructured CeO2 and SiC to ion irradiation are studied under different nuclear and electronic stopping powers to describe some general material response in this transit energy regime. Although fast radiation-induced grain growth in CeO2 is evident with no phase transformation, different fluence and dose dependence on the growth rate is observed under Si and Au irradiations. While grain shrinkage and amorphization are observed in the nano-engineered 3C SiC with a high-density of stacking faults embedded in nanosize columnar grains, significantly enhanced radiation resistance is attributed to stacking faults that promote efficient point defect annihilation. Moreover, competing effects of electronic and nuclear energy loss on the damage accumulation and annihilation are observed in crystalline 4H-SiC. Systematic experiments and simulation effort are needed to understand the competitive or synergistic effects.

Influence of oversized elements (Hf, Zr, Ti and Nb) on the thermal stability of vacancies in type 316L stainless steels

To reveal the influence of oversized elements on the thermal stability of vacancies in type 316L stainless steels, vacancy recovery processes were investigated by means of positron annihilation spectroscopy. Although vacancies in additive-free 316L stainless steels were mobile at 300°C, which is a typical nuclear reactor operating temperature, vacancies in oversized elements doped 316L were stable up to 300–350°C. This result indicates that oversized elements stabilize vacancies in stainless steels. Stability of vacancies inhibits the radiation-induced grain boundary segregation and may also lead to suppression of high-temperature water stress corrosion cracking that is observed in nuclear materials.

Irradiation Assisted Grain Boundary Segregation in Steels

AbstractThe understanding of radiation-induced grain boundary segregation (RIS) has considerably improved over the past decade. New models have been introduced and much effort has been devoted to obtaining comprehensive information on segregation from the literature. Analytical techniques have also improved so that chemical analysis of layers 1 nm thick is almost routine. This invited paper will review the major methods used currently for RIS prediction: namely, Rate Theory, Inverse Kirkendall, and Solute Drag approaches. A summary is made of the available data on phosphorus RIS in reactor pressure vessel (RPV) steels. This will be discussed in the light of the predictions of the various models in an effort to show which models are the most reliable and easy to use for forecasting P segregation behaviour in steels. A consequence of RIS in RPV steels is a radiation induced shift in the ductile to brittle transition temperature (DBTT). It will be shown how it is possible to relate radiation-induced P segregation levels to DBTT shift. Examples of this exercise will be given for RPV steels and for ferritic steels being considered for first wall fusion applications. Cr RIS in high alloy stainless steels and associated irradiation-assisted stress corrosion cracking (IASCC) will be briefly discussed.

A new model for radiation-induced grain boundary segregation with grain boundary movement in concentrated alloy system

We have developed a new model for radiation-induced grain boundary migration (RIGM) and radiation-induced segregation (RIS) for austenitic iron-chromium-nickel alloy system. It was assumed that the RIS was induced by diffusional and annihilation processes of excess point defects at the grain boundary, and the RIGM occurred due to rearrangement process of atoms on one of the interfacial planes by annihilation of point defects. The calculated results indicated that the region of RIS was enlarged by the RIGM and asymmetrical concentration profiles were observed around the migrated grain boundary. The present model could explain the RIS behavior with or without grain boundary migration as comparing with our previous experimental results.

Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking

Isolation of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) was attempted by means of low-dose high-temperature neutron irradiation in a material test reactor to get better understanding on IASCC. Microstructure, grain boundary segregation, hardness and SCC susceptibility were examined on stainless steels irradiated to 0.8 dpa at around 673 K. The irradiation caused well-developed grain boundary segregation without notable hardening or microstructural changes. No IASCC was found in 593 K hydrogenated water whereas the steels were highly susceptible to IASCC in 561 K oxygenated water. The results suggest that grain boundary segregation, probably Cr depletion, is sufficient to cause IASCC in oxygenated water and that other radiation-induced changes such as microstructure and hardening are required for IASCC in hydrogenated water.

Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking

Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking

Sink effect of grain boundary on radiation-induced segregation in austenitic stainless steel

We have investigated the orientation dependence of grain boundaries in radiation-induced segregation (RIS) in an austenitic stainless steel under irradiation using a high-voltage electron microscope (HVEM) at the Hokkaido University HVEM facility and a new rate equation model for RIS, which incorporates the grain boundary sink strength for point defects. The 〈110〉 tilt grain boundaries were chosen for the present study. It was observed that, after electron irradiation at a dose rate of 2.0×10−3dpas−1 to 10 dpa at temperatures of either 623 or 723 K, RIS was enhanced as the tilt angle increased but was suppressed at coincidence grain boundaries (Σ9 and Σ3). The present result indicates that one needs to explicitly consider the grain boundary sink strength as an important factor affecting radiation-induced grain boundary phenomena (RIGBPH). This work can be thought as a study of the radiation-induced grain boundary phenomena associated with grain boundary engineering related to non-equilibrium phenomena.

Effect of additional minor element on radiation-induced grain boundary segregation in austenitic stainless steel under electron irradiation

Radiation-induced segregation (RIS) influences the chemical and mechanical properties of structural materials of fission and fusion reactors. In the present study, we investigated the effect of minor alloying elements on RIS in austenitic stainless steel on the basis of the rate theory at or near grain boundaries. It was assumed that the additional minor elements interacted with vacancies and formed additive–vacancy complexes. The effects of additional element concentration and the binding energy between additives and vacancies on solute segregation at the grain boundary were clarified by the present calculations, and the results were compared with recent electron irradiation experiments using a high-voltage electron microscope (HVEM).

Effect of solute concentration on grain boundary migration with segregation in stainless steel and model alloys

The phenomenon of grain boundary migration due to boundary diffusion via vacancies is a well-known process for recrystallization and grain growth during annealing. This phenomenon is known as diffusion-induced grain boundary migration (DIGM) and has been recognized in various binary systems. On the other hand, grain boundary migration often occurs under irradiation. Furthermore, such radiation-induced grain boundary migration (RIGM) gives rise to solute segregation. In order to investigate the RIGM mechanism and the interaction between solutes and point defects during the migration, stainless steel and Ni–Si model alloys were electron-irradiated using a HVEM. RIGM was often observed in stainless steels during irradiation. The migration rate of boundary varied, and three stages of the migration were recognized. At lower temperatures, incubation periods up to the occurrence of the boundary migration were observed prior to first stage. These behaviors were recognized particularly for lower solute containing alloys. From the relation between the migration rates at stage I and inverse temperatures, activation energies for the boundary migration were estimated. In comparison to the activation energy without irradiation, these values were very low. This suggests that the RIGM is caused by the flow of mixed-dumbbells toward the grain boundary. The interaction between solute and point defects and the effective defect concentration generating segregation will be discussed.

Radiation-Induced Grain Boundary Segregation in Austenitic Stainless Steels

Radiation-induced segregation (RIS) to grain boundaries in Fe-Ni-Cr-Si stainless alloys has been measured as a function of irradiation temperature and dose. Heavy-ion irradiation was used to produce damage levels from 1 to 20 displacements per atom (dpa) at temperatures from 175 to 550{degrees}C. Measured Fe, Ni, and Cr segregation increased sharply with irradiation dose (from G to 5 dpa) and temperature (from 175 to about 350{degrees}C). However, grain boundary concentrations did not change significantly as dose or temperatures were further increased. Although interfacial compositions were similar, the width of radiation-induced enrichment or depletion profiles increased consistently with increasing dose or temperature. Impurity segregation (Si and P) was also measured, but only Si enrichment appeared to be radiation-induced. Grain boundary Si peaked at levels approaching 10 at% after irradiation doses to 10 dpa at an intermediate temperature of 325{degrees}C. No evidence of grain boundary silicide precipitation was detected after irradiation at any temperature. Equilibrium segregation of P was measured in the high-P alloys, but interfacial concentration did not increase with irradiation exposure. Comparisons to reported RIS in neutron-irradiated stainless steels revealed similar grain boundary compositional changes for both major alloying and impurity elements.

Radiation-Induced Grain Boundary Segregation

Experimental evidence for phosphorus segregation and chromium depletion at grain boundaries in steels is presented. Theories for irradiation-induced inter-granular segregation are described. Non-equilibrium (NES) and rate theory approaches have similar success in predicting phosphorus behaviour in the practically important temperature range although site competition and microstructural effects are better accounted for by the NES theory. NES theory predicts chromium behaviour reasonably well but there is a need to consider site competition more closely in relation to large atom effects from elements like Zr and Hf. The need for better data on diffusion constants and point defect-impurity binding energy is emphasised.

Effects of Interaction between Point Defects and Solutes on Grain Boundary Segregation and Migration

In order to investigate the mechanisms of radiation-induced grain boundary migration,a 316 model steel and Ni-(2-10)at%Si model alloys were electron-irradiated. Boundary migration was often observed in stainless steels and the model alloys during irradiation. An initial incubation and three migration stages were observed.The incubation period depended on irradiation temperature and solute concentration. From the temperature dependence of boundary migration velocity during the first stage and the incubation response, activation energies were estimated. The activation energies obtained for each alloy were lower than that for boundary migration under thermal annealing. It is suggested from these results that the grain boundary migration under irradiation is caused by enhanced boundary diffusion and preferential rearrangement of under sized interstitial solute atoms diffusing via a mixed dumbbell mechanism towards the grain boundary.

Radiation-induced grain boundary segregation in nuclear reactor steels

The paper will review the various mechanisms of grain boundary segregation. These are: thermal equilibrium segregation (TES); neutron irradiation enhanced equilibrium segregation; thermal non-equilibrium segregation; and radiation-induced segregation (RIS). The neutron irradiation-induced mechanisms will be discussed in detail. Modelling of the neutron irradiation segregation effects will be outlined with reference to two approaches: non-equilibrium theory based models and rate theory models. Both models have been successful in predicting phosphorus grain boundary segregation in pressure vessel steels as a function of irradiation conditions and examples of this success will be given in the paper. In addition the models have recently been adapted to account for the behaviour of Cr in austenitic steels with a view to understanding irradiation-assisted intergranular stress corrosion (IASCC). The use of predictive modelling of this kind in forecasting the embrittled state of grain boundaries as a function of reactor operating conditions will be emphasised. The influences of neutron dose, dose rate, temperature and microstructure will be mentioned.