Change in flow stress and ductility of δ-phase Pu-Ga alloys due to self-irradiation damage

An internal state variable model for the mechanical behavior of aged Pu-Ga alloys is developed and used to predict the change of the material with accumulated self-irradiation damage or age. The material model incorporates microstructural data such as the primary irradiation-induced defect density from cascades, the density and average size of helium bubbles, the initial dislocation density, and the initial average segment length of the dislocation density as input parameters, and then evaluates the stress-strain response of a representative volume element of the material. Given this response at a single material point, the deformation behavior of tensile specimens is predicted, and it forecasts increased strength, decreased strain hardening, and more strain localization with aging. Although the material point behavior showed some slight strain softening, this strain softening is expected to be masked by statistical variations of different volume elements and by the strain rate sensitivity of the material. Hence, it is not expected to appear in the stress-strain response of macroscopic tensile specimens, and only the increase in flow stress will be measured.     

Mechanical properties and microstructure of helium-implanted beryllium

0.2 mm thick specimens of beryllium have been homogeneously implanted with helium. Implantation temperatures ranged from 100 to 600 °C, and final helium concentrations from 30 to 800 appm. Tensile tests at temperatures between 20 and 600 °C were carried out with testing temperature both equal to and lower than the implantation temperatures. For practicality all conditions of helium-implanted specimens, ductility decreased and yield and ultimate tensile strength increased as compared to the unimplanted specimens. The amount of embrittlement and strengthening, however, depended sensitively on implantation dose, implantation temperature, and tensile test temperature. The formation of helium bubbles, dislocation loops, and dislocation networks and the fracture modes were observed by transmission and scanning electron microscopy, respectively. Two ranges of embrittlement can be distinguished. They are attributed to different mechanisms: matrix strengthening is the dominant mechanism at low temperatures, and loss of grain boundary cohesion at high temperatures. It is concluded that in both temperature regimes the embrittlement is dominated by helium and not by the displacement defects introduced by its implantation.     

Mechanical property degradation in irradiated materials: A multiscale modeling approach

High-energy particle irradiation of low stacking fault energy, face centered cubic (fcc) metals produces significant degradation of mechanical properties, as evidenced in tensile tests performed at or near room temperature. Post-irradiation microstructural examination reveals that approximately 90% of the radiation-induced defects in copper are stacking fault tetrahedra (SFT). Radiation damage is an inherently multiscale phenomenon involving processes spanning a wide range of length and time scales. Here we present a multiscale modeling methodology to study the formation and evolution of defect microstructure and the corresponding mechanical property changes under irradiation. At the atomic scale, molecular dynamics (MD) simulations have been used to study the evolution of high energy displacement cascades, SFT formation from vacancy rich regions of displacement cascades, and the interaction of SFTs with moving dislocations. Defect accumulation under irradiation is modeled over diffusional length and time scales by kinetic Monte Carlo (KMC), utilizing a database of displacement cascades generated by MD. The mechanical property changes of the irradiated material are modeled using three-dimensional dislocation dynamics (DD). Key input into the DD includes the spatial distribution of defects produced under irradiation, obtained from KMC, and the fate of dislocation interactions with SFTs, obtained from MD.     

ODS-Eurofer钢微观结构及辐照硬化研究

研究了ODS-Eurofer钢的微观结构及辐照硬化现象。首先用透射电子显微镜(TEM)观察了ODS-Eurofer钢的初始微观组织结构,发现基体中不仅存在几nm至几十nm的氧化物弥散颗粒,还存在具有壳 核结构的大尺寸(直径大于100 nm)颗粒,并观察到纳米颗粒对位错线的钉扎作用。随后用能量为5 MeV的Fe²⁺离子在300 ℃和500 ℃下辐照样品至25 dpa以模拟中子辐照,并用纳米压痕仪和TEM测试表征了辐照所致力学性能和微观结构的变化。结果表明,两种温度下辐照均引起硬度上升,500 ℃时由于辐照产生的点缺陷发生复合,导致硬化效应弱于300 ℃。用TEM观测辐照水平为25 dpa的损伤层发现有少量纳米尺寸位错环,这些位错环是辐照硬化的主要原因。ODS-Eurofer钢初始微观结构对辐照硬化有重要影响,其中晶界、纳米颗粒与基体界面、位错线等能捕获辐照过程中产生的点缺陷,从而抑制辐照位错环的生长。     

非均匀形核对辐照诱导钨内微结构演化行为影响的团簇动力学模拟

金属钨(W)及其合金作为未来聚变堆最具应用前景的面向等离子体结构材料(PFMs),其服役性能直接影响聚变堆长期服役的安全性,辐照诱导W及其合金内微结构演化导致的辐照脆化现象始终是限制其工程应用的关键因素。本文基于分子动力学计算结果,进一步完善了辐照诱导材料微结构演化行为的团簇动力学模型,采用更加完备的物理模型描述材料内辐照缺陷的产生行为,并进一步探讨了W基体内辐照缺陷产生过程对微结构演化行为的影响。模拟结果表明,高能初始离位原子(PKA)诱发级联碰撞直接产生的缺陷团簇是W内位错环、空洞演化中最重要的形核机制;非均匀形核所产生的间隙团簇的扩散行为对位错环的长大行为有重要影响,会导致位错环尺寸分布中出现亚尖峰与台阶状形貌。     

A polycrystalline modeling of the mechanical behavior of neutron irradiated zirconium alloys

Zirconium alloys used as fuel cladding tubes in the nuclear industry undergo important changes after neutron irradiation in the microstructure as well as in the mechanical properties. However, the effects of the specific post-irradiation deformation mechanisms on the mechanical behavior are not clearly understood and modeled. Based on experimental results it is discussed that the kinematic strain hardening is increased by the plastic strain localization inside the dislocation channels as well as the only basal slip activation observed for specific mechanical tests. From this analysis, the first polycrystalline model is developed for irradiated zirconium alloys, taking into account the irradiation induced hardening, the intra-granular softening as well as the intra-granular kinematic strain hardening due to the plastic strain localization inside the channels. This physically based model reproduces the mechanical behavior in agreement with the slip systems observed. In addition, this model reproduces the Bauschinger effect observed during low cycle fatigue as well as the cyclic strain softening.     

A generalized model for radiation-induced amorphization and crystallization of U3Si and U3Si2 and recrystallization of UO2

A rate-theory model of radiation-induced amorphization and crystallization of U₃Si during ion irradiation has been generalized to include U₃Si₂ and UO₂. The generalized model has been applied to ion-irradiation and in-reactor experiments on U₃Si and U₃Si₂ 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 UO₂ wherein the stored energy in the UO₂ 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 UO₂.     

Deformation mode maps for tensile deformation of neutron-irradiated structural alloys

The deformation microstructures of neutron-irradiated nuclear structural alloys, A533B steel, 316 stainless steel, and Zircaloy-4, have been investigated by tensile testing and transmission electron microscopy to map the extent of strain localization processes in plastic deformation. Miniature specimens with a thickness of 0.25 mm were irradiated to five levels of neutron dose in the range 0.0001-0.9 displacements per atom (dpa) at 65-100 °C and deformed at room temperature at a nominal strain rate of 10⁻³ s⁻¹. Four modes of deformation were identified, namely three-dimensional dislocation cell formation, planar dislocation activity, fine scale twinning, and dislocation channel deformation (DCD) in which the radiation damage structure has been swept away. The modes varied with material, dose, and strain level. These observations are used to construct the first strain-neutron fluence-deformation mode maps for the test materials. Overall, irradiation encourages planar deformation which is seen as a precursor to DCD and which contributes to changes in the tensile curve, particularly reduced work hardening and diminished uniform ductility. The fluence dependence of the increase in yield stress, ΔYS = α(?t)ⁿ had an exponent of 0.4-0.5 for fluences up to about 3 × 10²² n m⁻² (∼0.05 dpa) and 0.08-0.15 for higher fluences, consistent with estimated saturation in radiation damage microstructure but also concurrent with the acceleration of gross strain localization associated with DCD.     

高温无机闪烁晶体Ce:YAlO3的缺陷研究

采用同步辐射白光透射形貌术研究了提拉法生长的高温无机闪烁晶体Ce:YAlO3（简称Ce:YAP）中的缺陷。实验发现在Ce:YAP晶体中存在着生长条纹、包裹沉积物、核心、孪晶及位错簇等缺陷，同时对生长缺陷形成的原因进行了讨论。结果表明，离子掺杂浓度、原料的纯度以及生长工艺条件等是影响Ce:YAP晶体缺陷的主要原因。     

Correlation of radiation creep theory with experimental evidence

A review is presented of theoretical models related to irradiation creep and to the evolution of the dislocation structure in irradiated stainless steels. The results of detailed analysis for stress-induced loop alignment and stress-induced preferential absorption (SIPA) of point defects at dislocations is presented. Stress-induced rotation of tri-interstitials is shown to be too small to account for the observed variations in loop densities on different crystallographic planes. However, it is possible to predict large variations with the SIPA mechanism. Predictions of irradiation creep by the SIPA mechanism are in agreement with measured data at intermediate fluences. At low fluences, additional contributions to irradiation creep must come from dislocation glide. The evolution of the dislocation structure can be explained by the continuous formation of interstitial type loops and by the radiation-induced recovery of the dislocation network.     

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO2 fuel

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO₂ fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO₂ fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.     

Effects of edge dislocations on interstitial helium and helium cluster behavior in α-Fe

The properties of interstitial He in the vicinity of an edge dislocation were studied using molecular dynamics (MD) simulation. The distribution of the binding energy of a single interstitial He to the dislocation with and without a jog is calculated. The results show that the distribution of the binding energy is governed by the elastic interaction between the interstitial He and the dislocation. The interstitial He is strongly attracted to the dislocation in the tensile region of the dislocation. The jog acts as a stronger sink to absorb interstitial He. The binding energy to the jog is even larger than that of the dislocation. A small He cluster (composed of three interstitial He atoms) was trapped by the dislocation core in the form of a chain along the dislocation line. The dislocation changes the migration behavior of the He cluster, and provides a pipe for the small cluster to exhibit one-dimensional motion. The diffusion of the He cluster in the dislocation is faster than in the defect-free iron, where the He cluster migrates three-dimensionally (3D). If the dislocation is decorated by a jog, the small cluster sinks deep into the jog. The jog prevents the He cluster from escaping.     

辐照硬化位错动力学模拟的研究进展

辐照硬化是金属材料的辐照效应之一,开展辐照硬化机理研究有助于设计可靠的反应堆结构材料。辐照产生的缺陷会对位错运动造成阻碍,被认为是辐照硬化的主要原因。近年来快速发展的位错动力学模拟方法为材料的微观组织变化和宏观力学性能之间建立起了桥梁,被广泛用于辐照硬化机理研究。对于一些辐照缺陷如位错环和层错四面体,位错动力学软件已能模拟它们对位错网络演化以及宏观力学响应的影响,使辐照硬化的定量预测成为可能。本文从位错动力学模型、不同类型辐照缺陷硬化效应的位错动力学模拟以及辐照硬化理论模型发展三个方面,综述了辐照硬化位错动力学模拟的研究进展,并展望该研究领域的主要科学问题。     

Investigation of pulsed irradiation effects on interstitial loop growth in metals

A computer program for the solution of non steady-state diffusion equations describing the evolution of point defects and interstitial dislocation loops during pulsed and continuous irradiation is developed. The equations take into account mutual recombination of point defects, defect migration to dislocation loops and line dislocations, and the existence of equilibrium thermal vacancies. It is shown that interstitial loops grow from 2 to 9 run in diameter due to the surplus flux of interstitials in the non steady-state regime (dynamic preference) at 573 K. At 873 K the dislocation loops begin to shrink owing to line tension forces. Comparison of interstitial loop and vacancy behaviour for pulsed and continuous irradiation at 573 and 873 K is performed. It is shown that at pulse duration 2 × 10⁻⁶ s and repetition rate 100 pulses/s, pulsing does not affect the interstitial loop behaviour.     

Effects of an alloying element on a c-component loop formation and precipitate resolution in Zr alloys during ion irradiation

It is important to clarify the mechanisms of the dislocation loop formation, dissolution of precipitates to understand the degradation behavior of the fuel cladding tubes in light water reactors (LWR) under neutron irradiation. In this study, 3.2 MeV Ni ion irradiation was carried out at 400°C on Zircaloy-2 and two types of model alloys with and without Fe (Zr-1.5Sn-0.3Fe and Zr-1.5Sn). To understand the effects of hydrogen, 60 and 300 ppm pre-injected Zircaloy-2 samples were also irradiated. The microstructure was observed with a conventional transmission electron microscopy. Additionally, the dissolution of precipitates and the enrichment of the alloying element due to irradiation were analyzed using a spherical aberration (Cs)-corrected high-resolution analytical electron microscope. After ion irradiation at 400°C, the dissolution of Fe-enriched precipitates and the c-component dislocation loops were observed in the region of peak ion damage. Observations by STEM-EDS showed that Fe atoms were enriched in the c-component dislocation loops. On the contrary, the c-component dislocation loops were detected in Fe-containing alloys (Zircaloy-2 and Zr-1.5Sn-0.3Fe alloy) but were not in the Zr-1.5Sn alloy. These results indicate that the dissolution of Fe-enriched precipitates and the enhanced formation of c-component dislocation loops are essential for the degradation of LWR fuel cladding under irradiation.     

Failure and fracturing analysis of concrete structures

Some aspects of fracture analysis of concrete structures are discussed in this article. In particular it is shown that when localized failure occurs (by macrofracture propagation or localization of strain) structural size effects come into play. Mesh dependent finite element solutions are then observed unless size effects are correctly accounted for.Tensile fracture is examined first. The “classical” discrete and smeared crack approaches are reviewed and their extension to nonlinear fracture models like the fictitious crack model and the crack band model is illustrated. The smeared crack approach coupled first with a tensile strength criterion, second with a linear elastic fracture mechanics criterion is then applied to the failure mode analysis of a PCRV.Plastic fracturing with localization into shear bands, strain softening, mesh dependence and its correction are examined next. The use of plasticity for tensile fracture simulation is also discussed.Finally numerical difficulties inherent to the modeling of softening behavior are investigated.     

Effect of Neutron Irradiation on Deformation Behavior of Zirconium

The effect of neutron irradiation on the tensile deformation behavior of zirconium was examined at room temperature at various strain rates ranging of 2.2×10^?4~2.2× 10^?2 sec^?1. The microstructure of the deformed specimens was observed by transmission electron microscopy. It was established that neutron irradiation diminishes the uniform elongation and the strain hardening rate, and hastens the onset of plastic instability. These phenomena are attributed to inhomogeneous deformation in the dislocation channels in the irradiated and deformed zirconium.

From the relation between strain rate and tensile properties (yield stress, ultimate tensile stress, uniform elongation and strain hardening rate), it was established that in unirradiated zirconium deformation is controlled by slip at strain rates below 6×10^?3 sec^?1, while above this threshold, twinning as well as slip contribute to deformation.

Neutron irradiation markedly inhibits deformation twinning in zirconium at room temperature. At 77 K, on the other hand, deformation by twinning is more prominent in irradiated specimens. The mechanism of twinning inhibition due to neutron irradiation is discussed.     

Mechanisms of radiation induced creep and growth

Irradiation creep occurs primarily because the applied stress causes the evolving microstructure to respond in an anisotropic fashion to the interstitial and vacancy fluxes. On the other hand, irradiation growth requires the response to be naturally anisotropic in the absence of applied stress. Four fundamental mechanisms of irradiation creep have been conjectured: stress induced preferred absorption (SIPA) of the point defects on the dislocations, stress induced preferred nucleation (SIPN) of point defects in planar aggregates (edge dislocation loops), stress induced climb and glide (SICG) of the dislocation network and stress induced gas driven interstitial deposition (SIGD). These mechanisms will be briefly outlined and commented upon. The contributions made by these mechanisms to the total strain are not, in general, mutually separable and also depend on the prevailing (and changing) microstructure during irradiation. The fundamental mechanism of irradiation growth will be discussed: it is believed to arise by the preferred condensation of point defects and climb of dislocation loops and network on certain crystallographic planes. The preferred absorption and nucleation is thus a consequence of natural crystallographic anisotropy and not due to any external stresses. Again the effectiveness of this mechanism depends on the prevailing microstructure in the material. In this connection attention will be particularly drawn to the significance of solute trapping, segregation at grain boundaries, dislocation bias for interstitials and transport parameters for an understanding of irradiation growth in materials like zirconium and its alloys; the relevance of recent simulation studies of growth in such materials using electrons to the growth under neutron irradiation will be discussed in detail and a consistent model of growth in these materials will be presented.     

On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ∼7 × 10¹⁹ n cm⁻². This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.     

Modeling of the Ostwald ripening of extrinsic defects and transient enhanced diffusion in silicon

We present an atomistic simulation of the Ostwald ripening of extrinsic defects (clusters, {1 1 3}s and dislocation loops) which occurs during annealing of ion implanted silicon. The model describes the capture and emission of Si interstitial atoms to and from extrinsic defects of sizes up to thousands of atoms and includes a loss term due to the flux of interstitials to the recombining surface. Key input parameters of the simulation are the variations of the formation energy and of the capture efficiency with the size of the different defects. This model shows that the kinetics of the well-known dissolution of {1 1 3} defects is only driven by the recombination efficiency at the surface and the distance from the defects to the sample surface. We have subsequently used this model to study defect evolution in low and ultra low energy (ULE) B implanted Si during annealing. Defect dissolution occurs earlier and at smaller sizes in the ULE regime. Consequently, TED is mostly characterized by its “pulse” component which occurs at the very beginning of the anneal.