Void Number Density Research Articles

Characterization of Fe-Cr alloys irradiated by neutrons at intermediate temperature

A series of Fe-Cr alloys, with 3 at.% - 18 at.% Cr, was irradiated in the Advanced Test Reactor (ATR) at the Idaho National Labs (INL), USA, up to a dose of ∼6.7 dpa at a temperature of ∼456 °C. Transmission electron microscopy (TEM) samples were extracted using a focused ion beam (FIB) instrument, and the resulting microstructural defects, such as voids, dislocation loops, network dislocations, Cr rich precipitates, etc., were characterized using a TEM. It was found that the size and number density of these defects varied widely over the different alloys with varying Cr content. As expected, there were no Cr rich precipitates in samples with Cr up to 9 %, and they started appearing only in samples with 12 at.% Cr and above. The particle size decreased from about 15 nm at 12 % Cr to 8 nm at 18 % Cr, while the number density increased from ∼7e20 /m3 to 6e22 /m3 for the same Cr contents. Grain boundary segregation of Cr, along with a precipitate free zone, was observed in the cases where a boundary was present in the sample. Large voids (>1–2 nm) were almost invisible in the Fe-3at.%Cr sample, while the average void size remained almost constant between 15 and 18 nm for samples with 6–15 at.% Cr and increased slightly at 18 at.% Cr to ∼22 nm. Fe-3 %Cr showed a high density of small voids(<2 nm), estimated to be about 1e23/m3. The number density of large voids increased from ∼0 at 3 % Cr to a peak of ∼6.4e20 /m3 at 12 % Cr, then decreased to about 1.1e20 /m3 at 18 % Cr. The dislocation loops, which appeared in linear arrays in the Fe-9 %Cr sample, were analysed in detail using both invisibility criteria and image simulations using the Oxford University TEMACI software, and it was found that they are most likely to be ½ 〈111〉 type loops on {100} planes. These loops seem shaped like sections of helices in some places and are most likely formed by loops near screw dislocations climbing into a helix and then collapsing into a loop array. Similar dislocation analysis was performed in the other samples as well wherever feasible, and it was found that there is a mixture of ½ 〈111〉 and [100] dislocation loops.

Atomic-scale three-dimensional irradiation-induced defect kinetics models for bcc Fe-based alloys

The interaction between irradiation defects and dislocations could lead to macroscopic hardening and embrittlement. In this study, molecular dynamics (MD) and machine learning were conducted to study three-dimensional void- and Cu-cluster-induced kinetics models under interaction with dislocations in body-centered cubic Fe-based alloys. MD Results show that dislocation climb and shear are two types of interaction mechanism based on the defect size. Combined MD results and machine learning, it was found that the dislocation length increased linearly with an increase in the size of the defects after the interaction. In addition, the number of atoms in the Cu-rich cluster and the reduced number of vacancies in the voids had a square relationship with the defect size where 0.85D2 is for Cu-rich cluster and 0.50D2 is for void. Furthermore, atomic-scale three-dimensional irradiation-induced defect kinetics models were developed and incorporated into the crystal plasticity finite element model (CPFEM). We also analyzed the contributions of various defects to the increase in the yield stress during CPFEM simulation. Compared with DBH model, the prediction of MD data-driven CPFEM matches better with the experimental data. Cluster contributed the most to hardening with low obstacle strength due to the higher number density compared with dislocation loops and voids. The loops also contributed to hardening; however, their contribution was smaller than that of the clusters. Although the number density and size of voids were minimized, they may have contributed because of the high obstacle strengths. This work can indeed deepen the understanding of irradiation effect in Fe-based alloys.

Initial microstructure and temperature dependence of irradiation defects evolution in tungsten

Using the object kinetic Monte Carlo (OKMC) method and the sink strength model, we systematically investigated the influence of the initial microstructure and temperature on the aggregation, migration, removal, and resulting size distribution of defects emerging in tungsten (W) under neutron irradiation. It is found that the probability of defect absorption at grain boundaries is generally higher than that at dislocations. This suggests that the grain size (2 ∼ 20 μm) has a stronger effect on the evolution of irradiation defects compared to dislocation density (1010 m−2 ∼ 1013 m−2). As the grain size increases, the number of residual defects decreases but the size distribution of vacancies is basically unchanged. Besides initial microstructure, the evolution of radiation damage also depends on the irradiation temperature. The increase in temperature induces an initial increase in the number density of voids, and then followed by subsequent decreases, corresponding to the peak density at 800–1000 K. Furthermore, the defect evolution in irradiated W with different initial microstructures and temperatures is simulated, and the calculated results are compared with recent experiments. These results help us to understand the role of the initial microstructure of the material in the evolution of the irradiation defects.

The effect of thermal history on the additively manufactured AlSi10Mg alloy response to ion irradiation

Aluminum alloys, mainly from the 5xxx and 6xxx families, are used extensively in research reactors due to their suitable properties. However, alloys common in nuclear applications are not yet suitable for Additive Manufacturing (AM). The alloy AlSi10Mg has good engineering properties and is the most used and investigated AM Al-alloy. In this work, stress relieved (SR) samples of AM AlSi10Mg have been compared to samples that underwent a subsequent heat treatment at 500°C (SHT), producing two distinctly different microstructures. The SHT samples contain larger grains with a cleaner matrix and coarser silicon particles compared to the SR samples. Samples with the two microstructures were irradiated using silicon ions up to damage of ∼20 dpa at a depth of about 1 µm. To compare the response of both conditions to irradiation, properties were measured using nano-indentation and transient grating spectroscopy (TGS). Microstructure was characterized using TEM. Results show that the SHT condition is much more sensitive to irradiation, compared to SR. Irradiation caused thermal diffusivity to decrease by ∼5% in the SR condition and by more than 50% in the SHT condition. Hardness increased by 12 and 24% in the SR and SHT conditions, respectively. Also, the void number density was significantly larger in the SHT condition. No change was observed in young's modulus. The lower sensitivity of the SR condition is attributed to a higher sink density present in the initial microstructure, resulting in overall less irradiation damage retained in the material after irradiation. The explanation offered for the major drop in thermal diffusivity observed only in the SHT condition is the large size of the silicon particles, which increases their instability during irradiation, causing them to partly dissolve and release atomic silicon into the matrix around them. These silicon atoms are effective electron and phonon scatter centers. The findings of this work show that the pre-irradiated microstructure of AM AlSi10Mg has a crucial role in the material's performance in nuclear reactor core conditions. Careful thought and care need to be implemented to the heat treatment applied to an as-built part.

Voids and haloes-in-voids statistics as a probe of the expansion history of the Universe

ABSTRACT Structures in the Universe are arranged into the cosmic web. Distributions, statistics, and evolutions of the structures can be used as probes for cosmological models. We investigate the number density of voids and dark matter haloes-in-voids in the excursion set theory (EST). We study the Markov and non-Markov frameworks of EST in both spherical and ellipsoidal collapse models. Afterward, we compare the number density of voids and haloes-in-voids in the standard lambda cold dark matter and the reconstructed model. The reconstructed model is a model-independent reconstruction based on background observations. This work explores the effects of the collapse model barrier in the different EST frameworks on the statistics of voids and the statistics of haloes-in-voids. Finally, we find the hint that cosmological models can be distinguished by the number density of haloes-in-voids in the 1.0−2.5 redshift range. The maximum difference is observed in z ∼ 1.9.

Low-temperature tensile properties, deformation and fracture behaviors in the ferrite and austenite duplex stainless steel with various grain sizes

Ferrite and austenite duplex stainless steels with different grain sizes ranging from 2 to 11 μm were tensile deformed at room temperature and 77 K, and the effects of grain size on the tensile properties were discussed. The mechanism of improvement of the tensile properties by grain refinement was also discussed based on the observation results of the deformation and fracture behaviors. Specimens having identical phase ratios and chemical compositions in two phases but three different grain sizes ranging from 2 to 11 μm were prepared. At room temperature (RT), the strength increased, but the elongation decreased with decreasing grain size. At 77 K, both the strength and elongation were improved by grain refinement. In the specimen with the smallest grain size, a characteristic increase in the work-hardening rate after satisfying the plastic instability condition occurred during the tensile test at 77 K. Deformation-induced α′ martensite was formed during the tensile test at 77 K, and the volume fraction of α′ martensite increased as the grain size decreased from 11 to 2 μm. This resulted in the highest total strain at 77 K in the fine-grained specimen. The number density and size of voids were decreased by refining the grain size. The distance between two adjacent voids, calculated from the number density and size of voids, assuming a random distribution of voids, increased with grain refinement. It suppressed void coalescence and provided the highest true fracture stress in the fine-grained specimen. The Hall-Petch coefficient of the present duplex stainless steel became larger at 77 K than that at RT, resulting in a higher strength increase by grain refinement at 77 K. It can be concluded that the grain refinement in duplex stainless steel remarkably improves both strength and ductility at low temperatures owing to the high Hall-Petch coefficient, an increase in the volume fraction of deformation-induced α′ martensite, and a decrease in the number density of voids.

Void swelling in additively manufactured 316L stainless steel with hafnium composition gradient under self-ion irradiation

Compositionally-graded austenitic 316L stainless steel (SS) samples with five different Hafnium (Hf) concentrations (up to 1 wt.%) were additively manufactured by directed energy deposition and then were irradiated using 5 MeV Fe2+ ions to 50 displacements per atom (dpa) at 500, 550, and 600 °C, respectively. A rastering beam was used to ensure homogenous irradiations for the large areas of ∼1.40 cm2. Composition-dependent void evolution was evaluated. Void size, void number density, and void swelling all tend to decrease with increasing Hf concentration at all three temperatures. When the nominal Hf concentration increases to 1 wt.% in the doped additively manufactured (AM) 316L SS, the void swelling is over an order of magnitude lower than that in pure AM 316L. Atom probe tomography results showed that about 0.14 wt.% Hf dissolved in the matrix at the 1 wt.%. Hf nominal concentration. The suppression of void swelling by Hf addition shows qualitative agreement with the numerical calculation based on the vacancy trapping mechanism, indicating that the oversized Hf reduces the steady state vacancy concentration, and hence the incubation period of swelling is extended, resulting in a dramatic difference in void swelling. Other factors including grain boundaries, dislocation and dislocation cells, Hf-rich particles, and delta ferrite are secondary to this mechanism. This work shows that additive manufacturing enabled microalloying is promising for developing void swelling resistant materials.

Influence of Multi-Axial Stress on Damage Evolution and Change in Crystal Misorientation under Creep Loading of Hastelloy X

Ni based superalloy Hastelloy X is used for high temperature components of power generation gas turbines and aircraft jet engines. Creep damage occurs preferentially in stress concentration portions. Therefore, it is important to clarify creep damage evolution process and to develop creep damage detecting method under multiaxial with stress gradient for safety operation. In this study, creep tests have been conducted by using smooth and notch specimens which have multiple round notches with notch tip radius of 0.5 mm (R0.5) and 2.0 mm (R2.0) on a Hastelloy X. Crystal misorientation parameter GROD of damaged specimens was measured by an EBSD method. Creep rupture times of the notch specimens were longer than those of the smooth specimens showing notch strengthen effect. Larger amount of creep voids on grain boundaries was observed in the notch specimens than that in the smooth specimens. Void number density took the maximum value around the notch root in R0.5, and at the specimen center in R2.0 indicating that a distribution pattern depends on notch root shape. Those distribution patterns correspond to the maximum stress distribution thorough the notch root section. From the crystal misorientation measurement, GROD maps show that values of GROD increase with increasing creep damage, and that higher GROD values occurred at around notch root in the notch specimens. Change in GROD average at the notch root section corresponds to axial creep strain distribution obtained from creep analysis of the notch specimens. Eventually, an equation indicating relationship between axial creep strain and the normalized GROD average independent on temperature and stress states was derived.

Microstructural evolution in tungsten binary alloys under proton and self-ion irradiations at 800 °C

This study examined the effects of alloying elements such as Re and Ta on the microstructural evolution of recrystallized W under proton (1 MeV H+) and self-ion (18 MeV W6+) irradiations at 800 °C. Although the number density of voids increased with increasing proton-induced damage level, the void density in W-Re (R) and W-Ta (R) alloys were lower than that of pure W (R). Herein, the addition of Re and Ta to W suppresses the void formation process. In the proton-irradiated W-3%Re (R), a lot of dislocation loops were observed at 0.05 dpa which is the stage of nucleation. The evolution process up to 0.2 dpa was characterized by loop growth via the absorption of clusters and point defects. The dislocation loops then coalesce and grow large, and the dislocation lines become tangled at 1 dpa. At 0.05 dpa, the dislocation loops in pure W (R) have already evolved into the tangled dislocations. Solute Re may inhibit the mobility of small dislocation loops and SIA clusters. In W-3%Ta (R) irradiated at 0.05 and 0.2 dpa, the coalescence process of the elongated dislocation loops was observed. Solute Ta may inhibit the mobility of SIA clusters. Although no voids and rafts were observed in self-ion irradiated W-3%Re (R) to 0.2 dpa, not only dislocation loops but also voids and rafts were observed in pure W (R) to 0.2 dpa. Solute Re may inhibit the mobility of small dislocation loops and SIA clusters, so that solute Re would suppress the raft formation and then the void formation under self-ion irradiation. The dispersed barrier model was applied to the irradiation hardening of the proton-irradiated W-3%Re (R) up to 0.2 dpa before developing dislocation lines. The hardening in the proton-irradiated W-3%Re (R) up to 0.2 dpa is mainly due to TEM-visible voids and dislocation loops.

Electro-Chemo-Mechanical Failure of Solid Electrolytes Induced by Growth of Internal Lithium Filaments.

Growth of lithium (Li) filaments within solid electrolytes, leading to mechanical degradation of the electrolyte and even short circuit of the cell under high current density, is a great barrier to commercialization of solid-state Li-metal batteries. Understanding of this electro-chemo-mechanical phenomenon is hindered by the challenge of tracking local fields inside the solid electrolyte. Here, a multiphysics simulation aiming to investigate evolution of the mechanical failure of the solid electrolyte induced by the internal growth of Li is reported. Visualization of local stress, damage, and crack propagation within the solid electrolyte enables examination of factors dominating the degradation process, including the geometry, number, and size of Li filaments and voids in theelectrolyte. Relative damage induced by locally high stress is found to preferentially occur in the region of the electrolyte/Li interface having great fluctuations. A high number density of Li filaments or voids triggers integration of damage and crack networks by enhanced propagation. This model is built on coupling of mechanical and electrochemical processes for internal plating of Li, revealing evolution of multiphysical fields that can barely be captured by the state-of-the-art experimental techniques. Understanding mechanical degradation of solid electrolytes with the presence of Li filaments paves the way to design advanced solid electrolytes for future solid-state Li-metal batteries.

Effects of different silicon content on irradiation defects and hardening in 9Cr ferritic/martensitic steel

In order to explore the effect of silicon (Si) content on the irradiation behavior of ferritic/martensitic (F/M) steel, 9Cr F/M steel with Si of 0, 0.4%, 0.7%, and 1.0% in weight percent were prepared. All specimens were irradiated with 2.4 MeV iron ions at 550 °C to the fluence of 6.41 × 1016 ions cm−2, corresponding to a peak damage dose of 70 dpa. Transmission electron microscopy (TEM) was used to characterize the microstructure of the alloys before and after irradiation, and nanoindentation tests were used to assess the irradiation hardening. The addition of Si inhibited the growth of both M23C6 and MX precipitates and promoted the nucleation of MX precipitates, and the nucleation of the M23C6 precipitates was inhibited when the added Si was less than 0.7 wt.%, indicating that the precipitates can be regulated by Si content. The martensite lath width and the packet size of 0.4 wt.% Si content alloys are the largest and then decrease with the increase of Si content. As the Si content increases, the average size of dislocation loops decreases while the number density increases. Alloys without Si have the smallest void number density, followed by alloys with 0.7 wt.%, 0.4 wt.%, and 1.0 wt.% Si content. The hardness of the unirradiated alloy increased with the increase of Si content. After irradiation, the most severe radiation hardening occurred in the 0.4 wt.% Si alloy and the lowest in the 0.7 wt.% Si alloy.

A damage evolution model considering void shape effect and new damage level definition

Metal fracture under different stress states is one of the most concerned problems in engineering field. Damage evolution model is of great value to intensive comprehension on fracture mechanism and its engineering application. In this paper, considering that fracture surfaces are formed by void evolution, a new damage level with clear physical significance was defined as the ratio of the maximum statistical cross-sectional area of voids to the cross-sectional area of the representative elements. Combined with this new damage level, a novel damage evolution model was established based on void nucleation, uniform growth and shape change. Subsequently, two groups of quasi-static tensile tests were conducted. One group was called fitting group and used to fit numerical model parameters. Another one was called analysis group and chosen to investigate fracture and void evolution. DIC analyses and SEM tests were also carried out to analyze the influence of specimen shapes on strain distribution and void evolution. Moreover, the damage evolution model was applied to the numerical simulation to verify its reliability. The numerical results were basically consistent with the experimental results by comparing the force-displacement relationships, fracture strains, void number density distributions and void ellipticity distributions. The effects of specimen shapes on damage and void evolution were specifically investigated by numerical simulation. All results illustrated the proposed model provides a tool for predicting quasi-static ductile fracture under different stress states and has guiding value for engineering design.

Machine learning assisted investigation of defect influence on the mechanical properties of additively manufactured architected materials

Additive manufacturing techniques can introduce defects that worsen the mechanical properties of 3D printed parts. Current techniques for quantifying the detrimental effects of these defects can only provide detailed analysis for a small number of geometries. Here, we investigate the effect of each defect feature (surface roughness and void position, number density and size) on the mechanical properties of a large number of truss lattices belonging to the stretch-dominated and bending-dominated topology. This is done by reducing each truss lattice into a single-beam sub-unit cell and conducting finite element simulations on it. The generated data is subjected to machine learning algorithms to identify the most important defect and design features that determine the mechanical properties of the overall structure. Our results indicate that surface roughness, Rmax (i.e. peak-to-trough height), exceeding 10% of the beam diameter strongly reduces the specific modulus and strength of lattice structures, especially for bending-dominated geometries. Interior voids, on the other hand, adversely affect stretch-dominated geometries but improve the specific properties of bending-dominated structures by removing under-stressed material in the core of the beams and causing them to become more “tube-like”. These insights are supported by first-principles analytical modeling and experimental data of additively manufactured metal lattices in the literature.

Evolution of microstructure in neutron irradiated cold rolled tungsten and its correlation with hardness

The understanding of the neutron irradiation effects in tungsten is of significant importance for its use as a plasma facing material in future fusion devices. In this study, cold rolled tungsten is neutron irradiated at the Belgian BR2 fission reactor at the temperatures 600, 800, 900 and 1200 °C to a dose of 0.18 displacements per atom (dpa). The neutron induced changes in the microstructure are investigated as a function of irradiation temperature using transmission electron microscopy, positron annihilation spectroscopy and electrical resistivity measurements. The influence of the irradiation on the elastic properties and the hardness is examined using the impulse excitation techniques and depth-sensing indentation. Voids and dislocation loops are observed at all the irradiation temperatures. As the irradiation temperature increases the number density of both voids and loops decreases whereas their size increases. The total dislocation density increases after irradiation at 600 °C whereas it decreases for higher temperature irradiations. Furthermore, the formation of very small vacancy clusters in the temperature range of 800 - 900 °C is revealed by PAS measurements. A systematic decrease in the values of Young's and shear moduli is observed as the irradiation temperature increases and this decrease is of about 3.5% after irradiation at 1200 °C. A considerable hardening effect is observed at all irradiation temperatures. The hardness increases with irradiation temperature reaching a maximum at 800 °C and remains almost constant at higher irradiation temperatures.

The Modified Void Nucleation and Growth Model (MNAG) for Damage Evolution in BCC Ta

A void coalescence term was proposed as an addition to the original void nucleation and growth (NAG) model to accurately describe void evolution under dynamic loading. The new model, termed as modified void nucleation and growth model (MNAG model), incorporated analytic equations to explicitly account for the evolution of the void number density and the void volume fraction (damage) during void nucleation, growth, as well as the coalescence stage. The parameters in the MNAG model were fitted to molecular dynamics (MD) shock data for single-crystal and nanocrystalline Ta, and the corresponding nucleation, growth, and coalescence rates were extracted. The results suggested that void nucleation, growth, and coalescence rates were dependent on the orientation as well as grain size. Compared to other models, such as NAG, Cocks–Ashby, Tepla, and Tonks, which were only able to reproduce early or later stage damage evolution, the MNAG model was able to reproduce all stages associated with nucleation, growth, and coalescence. The MNAG model could provide the basis for hydrodynamic simulations to improve the fidelity of the damage nucleation and evolution in 3-D microstructures.

Dose rate effects on damage accumulation and void growth in self-ion irradiated tungsten

• Disordering rates at the damage peak in tungsten (W) irradiated to 1 dpa at 900 K is ~6 times lower at 10 −4 than 10 −3 dpa/s. • A uniform distribution of voids is observed in the self-ion irradiated W at both dose rates. • The void diameter varies from 1.5 to 3.4 nm with the number density on the order of 2 × 10 17 voids/cm 3 . • Void growth in W irradiated at 900 K is faster at 10 −4 than 10 −3 dpa/s. • Void growth is faster in monocrystalline than polycrystalline W irradiated at 900 K. • A simplified three-dimensional model is proposed to assess the dose rate effects. This study reports on dose rate effects in self-ion irradiated monocrystalline tungsten (mono-W) and polycrystalline tungsten (poly-W). Both mono-W and poly-W were irradiated at 900 K to 1 dpa at two dose rates, 10 −4 and 10 −3 dpa/s. An additional mono-W disk was irradiated at 300 K up to 44 dpa. The irradiated samples were analyzed using Rutherford backscattering spectrometry in channeling geometry (RBS/C) and transmission electron microscopy. The RBS/C data suggest that self-interstitial atoms in mono-W are extremely mobile at 300 K and higher; the disordering rate in mono-W irradiated at 900 K is ~6 times lower at 10 −4 than 10 −3 dpa/s at the depth of the damage peak. There is a uniform distribution of voids in both mono-W and poly-W irradiated at 900 K. The observed void diameters were corrected, which ranged from 1.5 to 3.4 nm. The void number density is on the order of 2 × 10 17 voids/cm 3 in mono-W irradiated at 900 K to 1 dpa at 10 −3 dpa/s. The growth rate of the voids in W irradiated at 900 K is higher at 10 −4 than 10 −3 dpa/s. At the same dose rate, voids grow faster in mono-W than poly-W. A simplified three-dimensional model is proposed to assess the dose rate effects based on the diffusion and interaction of migrating point defects from two consecutive damage cascades.

Creep Damage Process and Rupture Time Prediction Methods of 2.25Cr-1Mo Steel Round Notch Bar Specimens

Creep damage preferentially progresses at stress concentration portion, where stress gradient exists under multiaxial stress state, in high temperature components. Therefore it is necessary to clarify creep damage evolution process and to establish a rupture time prediction method under multiaxial stress state. In this study, creep tests using plane specimens and four kinds of round notch bar specimens with different notch radius (notch tip radius 0.1mm(R0.1), 0.5mm(R0.5), 2.0mm(R2.0) and 4.0mm(R4.0)) on a 2.25Cr-1Mo steel were carried out. Rupture times of notch specimens were much longer than those of plane specimens under the same nominal stress indicating notch strengthening effect. Measured void number density on the notch root section indicated higher value at around notch surface for R0.1 and R0.5 specimens, while it was higher at around specimen center for R2.0 and R4.0. These distribution patterns corresponded to those of the maximum stress and the triaxiality factor of the notch specimens. Although, creep rupture times of notch specimens were not able to be predicted by using the stress or strain at the representative point on the notch root section, they were well predicted based on the area average damage concept, especially by considering selected damage area for R0.1 and R0.5 specimens.

Effect of Cooling Rates from Solution Treatment Temperature on Aging Response and Fracture Behavior of a Mg–8Al–0.5Zn Alloy

In the present study, a Mg–8Al–0.5Zn (AZ80) Mg alloy was subjected to water quenching and furnace cooling from the solution treatment temperature of 390 °C. Subsequently, all the samples were subjected to aging treatment at 250 °C. The precipitation behavior, aging response and fracture phenomenon of the AZ80 Mg alloy were evaluated to establish the correlation among the cooling rates after solution treatment, microstructural features and mechanical properties. Scanning electron microscopy depicted a very distinct precipitation behavior of aged specimens. The observed variation in the aging curves was rationalized based on this. Fractographic analysis on the as-solutionized and aged alloys revealed that the initiation of fracture (i.e., intergranular or transgranular) was dictated by the presence of microstructural features, such as precipitates (continuous and discontinuous), twins and grain boundaries. The fracture strain was correlated with the number density and size of voids formed during tensile loading.

Image processing based automated solution for analysis of dimple fracture surfaces

The present work aims to develop an automated solution for analysis of dimple tensile fracture surfaces. An image processing based technique has been proposed to extract the voids from the fractographic images of dimple fracture surfaces and also to analyse the void features. A two-stage process has been developed to extract the closed contour of the voids. At the first stage a global thresholding is followed by a sequence of processing, like thinning and morphological closing. In the second stage refinement process has been carried out on the large voids based on local characteristics. The proposed technique has been applied on the scanning electron fractographs of tensile fracture surfaces obtained by tensile testing of austenitic stainless steel at four different strain rates. It has been found that the size and density of extracted voids bear systematic relationship with tensile properties of the investigated steel. The average circular diameter with yield strength and tensile strength follows linear relationship. The number density of large voids is found to decrease with increase of strain rate and it also decreases with increase of average circular diameter in a linear manner. Similar nature of variation of tensile ductility has been observed both with strain rate and average circular diameter.

Creep damage analysis of mod.9Cr-1Mo steel welds considering void mechanics modeling

Since creep strength reduction of mod.9Cr-1Mo steel welds in long-term creep was confirmed experimentally, residual life prediction for Type IV creep failure has been an important theme in Ultra Super Critical (USC) plants after long-term service. The mechanism of Type IV creep failure within the fine-grained heat affected zone (FGHAZ) is twofold; an increase in number density of creep voids of a size comparable to FGHAZ grain size (5 μm), and strain softening in the last stage of creep life. In this paper, a creep damage analysis scheme considering the above two mechanisms is applied to welded joints with non-uniform weld metal materials, heat affected zone (HAZ) and base metal, such as those found in large uniaxial cross-weld specimens. A void mechanics model expresses the increase of number density of creep voids (creep void density) in terms of a strain for void nucleation, which is strongly influenced by the multiaxial state of stress. A critical value of creep void density corresponding to the initiation of a micro-crack is determined based on the results of a random-fracture-resistance model of grain boundaries by the authors of this paper. Both the creep void density distribution and the failure process with crack initiation, propagation and final rupture predicted by the analysis are compared with experimental results at 650 °C．