Irradiation-induced Recrystallization Research Articles

Neutron irradiation-induced recrystallization simulation for tungsten in nuclear fusion device

In magnetic confinement nuclear fusion devices, Tungsten (W) stands out as the most promising material for plasma-facing components, especially for divertor targets exposed to high neutron loads and high heat flux. However, the harsh operational conditions subject W undergoes significant degradation due to irradiation-induced defect production and irradiation-induced recrystallization, resulting in a modification of the original microstructure. These irradiation effects lead to a consequential deterioration in mechanical properties and component durability.This study implements a neutron irradiation-induced recrystallization (NIIR) modeling framework to quantify the recrystallization fraction in the W section of the plasma-facing components (PFCs). The simulation results reveal that the rate of neutron irradiation damage significantly impacts NIIR, leading to varying durations for achieving complete recrystallization in W at elevated temperatures. Furthermore, the half recrystallization time of W experiences a substantial reduction under neutron irradiation, resulting in a thicker recrystallized layer in W-PFCs compared to those without neutron irradiation but at the equivalent heat flux conditions.

Simulation of neutron irradiation-induced recrystallization of tungsten

<sec>Tungsten is the candidate for divertor target material in future fusion reactors. The tungsten divertor target is expected to long serve in a harsh environment of high temperature and high-energy neutron irradiation. This can lead to neutron irradiation-induced recrystallization of tungsten, thereby increasing the possibility of intergranular brittle failure and compromising the safe operation of the divertor. Thus, clarifying the mechanism of neutron irradiation-induced tungsten recrystallization is important. However, the current model, which only considers the irradiation-enhanced effect on recrystallization driving force, underestimates the irradiation effect on recrystallization compared with the results observed in recent high-temperature neutron irradiation experiments in the HFIR reactor. It indicates that other irradiation effects can also influence the recrystallization process.</sec><sec>In this study, we introduce the irradiation-enhanced grain boundary migration factor (<i>R</i>) into the established irradiation-induced recrystallization kinetic model, on the assumption that the grain boundary migration velocity is proportional to the self-diffusion coefficient. The simulation results show that after considering both irradiation-enhanced recrystallization driving force and grain boundary migration effect, the calculated half-recrystallization time (<inline-formula><tex-math id="M4">\begin{document}${t}_{{X}\text{}=\text{}0.5}$\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.png"/></alternatives></inline-formula>) at 850 ℃ from the model matches the one obtained in the neutron irradiation experiment in the HFIR reactor. This result indicates that the irradiation-enhanced grain boundary migration effect is one of the important factors affecting irradiation-induced recrystallization. In addition, the difference between irradiated and unirradiated <i>t</i><sub><i>X</i>=0.5</sub> decreases with temperature increasing. This phenomenon is due to the fact that as the temperature increases, the contribution of irradiation defects to the driving force for recrystallization decreases owing to the irradiation defect recombination. Moreover, the increase of thermal activation diffusion coefficient is more significant than the increase of the irradiation-enhanced diffusion coefficient. These findings suggest that the thermal activation effect eventually dominates the recrystallization process over the irradiation effect as temperature increases.</sec>

Influence of defect dynamics on the nanoindentation hardness in NiCoCrFePd high entropy alloy under high dose Xe+3 irradiation

We report the role of irradiation-induced defects and microstructure evolution in the wake of 1.05 MeV Xe3+ ion irradiation on the structural and mechanical properties of single-phase face-centered cubic (FCC) structured NiCoCrFePd high-entropy alloy (HEA). The defect evolution was investigated using positron annihilation Doppler broadening spectroscopy (PADBS), X-ray diffraction (XRD), high-resolution transmission electron microscopy (HR-TEM), and nanoindentation techniques. Single-phase FCC structured NiCoCrFePd HEA remained structurally stable even upon irradiation to a very high fluence of 9 × 1016 ions/cm2. However, an anomalous reduction in the lattice constant, along with micro-strain relaxation and crystallite fragmentation was observed at the initial ion fluence of 1 × 1016 ions/cm2, which are recovered upon irradiation at the higher ion fluence due to collapse of small-sized point defects into the large-sized dislocations and simultaneously irradiation-induced recrystallization. Ion fluence-dependent PADBS analysis revealed the formation of mono-vacancies at lower fluence which were saturated by recombination and evolution to large-size defects at successive higher ion fluence that is consistent with electron microscopic investigations. Measurement of mechanical properties showed that the hardness was initially increased by 57% upon irradiation at fluence 1 × 1016 ions/cm2 and then it was slightly reduced (∼5%) at a higher fluence of 9 × 1016 ions/cm2 irradiation due to defect recombination and recrystallization which clearly shows the resistance of this HEA towards hardening. Thus, the present study provides a deeper understanding of the defect dynamics and their relation to the mechanical behavior of NiCoCrFePd HEA, which is important for the development of radiation-resistant alloys for nuclear energy systems.

Radiation-induced amorphization and recrystallization of hydroxyapatite nanoparticles

Hydroxyapatite, Ca10(PO4)6(OH)2, is considered as an important apatite-type material for the incorporation and disposal of actinides and fission products. Hydroxyapatite nanoparticles with different size ranging from 20 nm to 280 nm are synthesized via calcining the bovine bones under different temperatures and durations. The samples are irradiated with 1 MeV Kr2+ ions and 200 keV electrons to study the displacive and ionizing effects on the irradiation behaviors. In situ transmission electron microscopy (TEM) observation shows hydroxyapatite nanoparticles can be amorphized by 1 MeV Kr2+ ions and the previously amorphized samples experience a rapid ionizing-radiation-induced recrystallization upon 200 keV electrons irradiations. A strong size effect on the displacive irradiation-induced amorphization and ionizing irradiation-induced recrystallization are observed. Under ion irradiation, a lower critical temperature (Tc) is observed for a larger sized sample, indicating enhanced amorphization tolerance. This result indicates excess interface energy resulting from the smaller sized hydroxyapatite nanoparticle, may lower the energy difference between the crystalline phase and amorphous state, degrading the radiation tolerance. Radiation-induced amorphization is attributed to the accumulation of oxygen vacancies upon ion irradiation and the glass-like structure formed by the rearrangement of displaced oxygen and phosphorus atoms. Whereas under electron irradiation, the recrystallization fluences decrease with the reduction of particle size since a higher density of dangling bonds exist in a smaller sized hydroxyapatite. These unstable bonds are relatively easy to break and reforming under ionizing electron irradiation, which drives the recrystallization process.

<i>In situ</i> TEM Study of Size Dependence on Radiation-Induced Amorphization and Recrystallization of Hydroxyapatite Nanoparticles

Hydroxyapatite, Ca10(PO4) 6(OH)2, is considered as an important apatite-type material for the incorporation and disposal of actinides and fission products. Hydroxyapatite nanoparticles with different size ranging from 20 nm to 280 nm were synthesized via calcining the bovine bones under different temperatures and durations. The samples are irradiated with 1 MeV Kr2+ ions and 200 keV electrons to study the displacive and ionizing effects on the irradiation behaviors. In situ transmission electron microscopy (TEM) observation shows hydroxyapatite nanoparticles can be amorphized by 1 MeV Kr2+ ions and the previously amorphized samples experience a rapid ionizing-radiation-induced recrystallization upon 200 keV electrons irradiation. A strong size dependence on the displacive irradiation-induced amorphization and ionizing irradiation-induced recrystallization are observed. Under ion irradiation, a lower critical temperature T c is observed for a larger sized sample, indicating enhanced amorphization tolerance. This result indicates excess surface energy as a result of larger surface area resulting from the reduced size, may lower the energy barrier between the crystalline phase and amorphous state, degrading the radiation tolerance. Whereas under electron irradiation, the recrystallization fluences decrease with the reduction of particle size since a higher density of dangling bonds exist in a smaller sized hydroxyapatite. These unstable bonds are relatively easy to break and reforming under ionizing electron irradiation, which drives the recrystallization process.

Suppressed radiation-induced dynamic recrystallization in CrFeCoNiCu high-entropy alloy

We report a map of irradiation-induced recrystallization and polygonization in a CrFeCoNiCu high-entropy alloy obtained by a precession electron diffraction technique. This orientation map enables the visualization of depth-dependent recrystallization at a glance and reveals the effects of irradiation dose and material entropy. Discontinuous dynamic recrystallization preferentially occurs near the surface, where the irradiation dose is low, but radiation-enhanced diffusion dominantly occurs leading to dislocation climb even at room temperature. Interestingly, in the high-entropy phase with a relatively low stacking fault energy, discontinuous dynamic recrystallizastion is suppressed due to the lower stored energy and grain boundary diffusivity compared low-entropy one.

Defect cluster and nonequilibrium gas bubble associated growth in irradiated UMo fuels – A cluster dynamics and phase field model

Irradiation examination shows that gas bubble swelling kinetics are much faster after irradiation-induced recrystallization than that prior to recrystallization in U-10 wt% Mo alloy (UMo) fuels. These kinetics imply that gas bubbles in coarse grains and small recrystallized grains have different growth behavior. For the first time, researchers developed a phase-field model of gas bubble evolution integrating microstructure-dependent cluster dynamics to study the gas bubble swelling behavior in the recrystallization zone of UMo fuels. Generation, diffusion, reaction, sink, emission, and clustering of vacancies and interstitials are described by the cluster dynamics model while a phase-field model is used to describe the evolution of nonequilibrium gas bubbles including nucleation and growth. With the coupled model, the effect of defect generation rate, clustering rate, interstitial emission, and sink rates on grain boundaries on the gas bubble evolution are systematically simulated. A set of model parameters (defect generation rate, clustering rate, interstitial emission, and sink rates) is determined by comparing measured and simulated gas bubble swelling kinetics. The results demonstrate that interstitial clustering is one of the important physical mechanisms that results in fast gas bubble swelling kinetics in the recrystallization zone. The developed model can also be extended to study the associated growth of defect and second-phase precipitates often observed in irradiated materials.

A Monte Carlo model of irradiation-induced recrystallization in polycrystalline UMo fuels

Experiments show that irradiation-induced recrystallization speeds up the swelling kinetics in U-10 wt% Mo fuels. However, recrystallization mechanisms and the effect of initial grain microstructures on recrystallization kinetics are still unclear. In this work a Monte Carlo model coupling the rate theory of defect evolution has been developed to study the irradiation-induced recrystallization. The rate theory is used to describe the spatial evolution of gas bubbles, interstitials and interstitial loops; First-Passage Kinetic Monte Carlo (FPKMC) approach is used to describe the fast and strongly anisotropic migration of interstitials, and a Cellular Automata method is used to model the formation of recrystallized grains. With the assumption that recrystallization may occur when the local interstitial loop density is larger than a given critical value, simulation results reveal that 1) recrystallized grains first nucleate on grain boundaries and the recrystallization zone front moves to the center of original coarse grains in the UMo matrix, and 2) recrystallization starts earlier in coarse polycrystalline structures, while the overall recrystallization kinetics decreases with increasing grain size. These results agree with experimental observations. The comparison of recrystallization kinetics obtained from experiments and modeling suggests that the interstitial loop accumulation leads to the recrystallization and the interstitial loop growth is suppressed inside coarse grains due to the over-pressured intra-granular gas bubbles.

Evolution of fission-gas-bubble-size distribution in recrystallized U–10Mo nuclear fuel

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles, used to characterize fission-gas bubble development in U–Mo alloy fuel with burnup limited to less than 10 at.% U in order to capture the fuel swelling stage prior to irradiation-induced recrystallization, is extended to recrystallized fuel at a burnup of ∼16 at.% U. During recrystallization the grain size is transformed from micron to sub-micron sizes. The intergranular bubble-size distribution post-recrystallization is found to evolve with similar kinetics and morphology to that pre-recrystallization with any differences primarily due to gas content and initial and/or boundary conditions (e.g., fuel microstructure). The predictions of the theory are compared with measured bubble-size distributions in pre and post recrystallized U–10Mo alloy fuel.

Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium–molybdenum alloy fuel

An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U–Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ∼7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U–Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15–125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

Derivation of analytical expressions for the network dislocation density, change in lattice parameter, and for the recrystallized grain size in nuclear fuels

At relatively high burn-up, both UO 2 and U–10Mo nuclear fuels undergo irradiation-induced recrystallization wherein grains of micron size transform to grains of submicron size. Previously, expressions have been derived for the initiation, and for the progression of recrystallization as a function of fission density, fission rate, fuel temperature, and as-fabricated grain size. Here, analytical expressions are derived for the network dislocation density, change in lattice parameter, and the size of the recrystallized grains. The basic premise of the theory presented in this work is that irradiation-induced recrystallization is driven by the behavior of interstitial loops. It is demonstrated that these phenomena can be simulated in both UO 2 and in U– xMo with the same theory, albeit with various property differences. Results of the calculations are compared with available data.

A model for the effect of the progression of irradiation-induced recrystallization from initiation to completion on swelling of UO 2 and U–10Mo nuclear fuels

Analytical expressions are derived for the progression of recrystallization from initiation to completion, and for the swelling due to fission-gas. It is demonstrated that these phenomena can be simulated in both UO 2 and in U– xMo with the same theory, albeit with various property differences. Results of the calculations are compared with available data.

05/00622 A model for the influence of microstructure, precipitate pinning and fission gas behavior on irradiation-induced recrystallization of nuclear fuels

05/00622 A model for the influence of microstructure, precipitate pinning and fission gas behavior on irradiation-induced recrystallization of nuclear fuels

A model for the influence of microstructure, precipitate pinning and fission gas behavior on irradiation-induced recrystallization of nuclear fuels

Irradiation-induced recrystallization appears to be a general phenomenon in that it is observed to occur in a variety of nuclear fuel types, e.g. U–xMo, UO2, and U3O8. For temperatures below that where significant thermal annealing of defects occurs, an expression is derived for the fission density at which irradiation-induced recrystallization is initiated that is athermal and weakly dependent on fission rate. The initiation of recrystallization is to be distinguished from the subsequent progression and eventual consumption of the original fuel grain. The formulation takes into account the observed microstructural evolution of the fuel, the role of precipitate pinning and fission gas bubbles, and the triggering event for recrystallization. The calculated dislocation density, fission gas bubble-size distribution, and fission density at which recrystallization first appears are compared to measured quantities.

The effect of irradiation-induced gas-atom re-solution on grain-boundary bubble growth

The rim region of high-burnup fuels is characterized by an exponential growth of intergranular porosity. In particular, the understanding of the dynamics of irradiation-induced recrystallization and subsequent gas-bubble swelling requires a quantitative assessment of the nucleation and growth of grain-boundary bubbles. Calculations of bubble growth on the grain boundaries of irradiated nuclear fuels at relatively low temperatures have, in general, been performed under the assumption that these bubbles are not appreciably affected by irradiation-induced gas-atom re-solution. In contrast, matrix bubbles are strongly affected by this bubble shrinkage mechanism and as a consequence are generally two to three orders or more of magnitude smaller than the grain-boundary bubbles. A variational method is used to calculate diffusion from a spherical fuel grain. The junction position of two trial functions is set equal to the bubble gas-atom knock out distance. The effect of grain size, gas-atom re-solution rate and diffusivity, gas-atom knock out distance, and grain-boundary bubble density on the growth of intergranular bubbles is studied, and the conditions under which intergranular bubble growth occurs are elucidated.

Irradiation-Induced Recrystallization of Cellular Dislocation Networks in Uranium-Molybdenum Alloys

ABSTRACTWe developed a rate-theory-based model to investigate the nucleation and growth of interstitial loops and cavities during low-temperature in-reactor irradiation of uranium-molybdenum alloys. Consolidation of the dislocation structure takes into account the generation of forest dislocations and capture of interstitial dislocation loops. The theoretical description includes stress-induced glide of dislocation loops and accumulation of dislocations on cell walls. The loops accumulate and ultimately evolve into a low-energy cellular dislocation structure. Calculations indicate that nanometer-size bubbles are associated with the walls of the cellular dislocation structure. The accumulation of interstitial loops within the cells and of dislocations on the cell walls leads to increasing values for the rotation (misfit) of the cell wall into a subgrain boundary and a change in the lattice parameter as a function of dose. Subsequently, increasing values for the stored energy in the material are shown to be sufficient for the material to undergo recrystallization. Results of the calculations are compared with SEM photomicrographs of irradiated U- 10Mo, as well as with data from irradiated UO2.

Application of a mechanistic model for radiation-induced amorphization and crystallization of uranium silicide to recrystallization of UO 2

An alternative mechanism for the evolution of recrystallization nuclei is described for a model of irradiation-induced recrystallization of UO 2 wherein the stored energy in the UO 2 is concentrated in a network of sinklike nuclei that heretofore were assumed to diminish with dose due to interaction with radiation-produced defects. The sinklike nuclei are identified as cellular dislocation structures that evolve relatively early in the irradiation period. In the alternative approach, a generalized theory of radiation-induced amorphization and crystallization, developed for uranium silicide, is applied to UO 2. The complicated kinetics involved in the formation of a cellular dislocation network are approximated by the formation and growth of subgrains due to the interaction of shock waves produced by fission-induced damage to the UO 2.

A generalized model for radiation-induced amorphization and crystallization of U 3Si and U 3Si 2 and recrystallization of UO 2

A rate-theory model of radiation-induced amorphization and crystallization of U 3Si during ion irradiation has been generalized to include U 3Si 2 and UO 2. The generalized model has been applied to ion-irradiation and in-reactor experiments on U 3Si and U 3Si 2 and provides an interpretation for the amorphization curve (dose required to amorphize the material as a function of temperature), for the ion-radiation-induced nanoscale polycrystallization of these materials at temperatures above the critical temperature for amorphization, as well as for the role of the small crystallites in retarding amorphization. An alternative mechanism for the evolution of recrystallization nuclei is described for a model of irradiation-induced recrystallization of UO 2 wherein the stored energy in the UO 2 is concentrated in a network of sinklike nuclei that diminish with dose due to interaction with radiation-produced defects. The sinklike nuclei are identified as cellular dislocation structures that evolve relatively early in the irradiation period. The complicated kinetics involved in the formation of a cellular dislocation network are approximated by the formation and growth of subgrains due to the interaction of shock waves produced by fission-induced damage to the UO 2.

Irradiation-induced recrystallization in high burnup UO2 fuel

The formation mechanism of recrystallized grains and coarsened bubbles in the peripheral region of a high burnup UO2 pellet has been proposed on the basis of transmission electron microscopy (TEM) results. TEM observations were made on the pellet edge (local burnup: 100 GWd/t) of a fuel sample which had been irradiated to a pellet average burnup of 49 GWd/t in a BWR. A recrystallized grain region was found adjacent to an accumulation region having an extremely high density of dislocations. The dislocations were organized into subdivided grains with high angle boundaries which were regarded as the nuclei for recrystallization. Almost all the coarsened bubbles were surrounded by recrystallized grains. Small intergranular bubbles precipitated on the recrystallized grain boundaries and interconnected with each other, and then linked up with the coarsened bubbles. From these results, it was concluded that the coarsened bubbles were formed by sweeping out of small bubbles during grain growth on recrystallization. Finally an overall formation process for the unique microstructural change, the rim structure, was proposed.