Irradiation Hardening Behavior Research Articles

Damage behavior of additive manufacturing 316L stainless steel irradiated with Fe ions

The irradiation damage behavior of 316L stainless steel (SS) fabricated using selective laser melting (SLM) and traditional manufacturing (TM) was investigated in terms of microstructural changes in hardness and swelling at RT under irradiation with 3 MeV Fe ions at doses of 0.3 to 6 dpa. At a dose of 6 dpa, the swelling heights of the SLM and TM samples were 39.1 nm and 30.7 nm, respectively, corresponding to swelling rates of 3.0 % and 2.4 %. The irradiated SLM 316L SS exhibited a lower degree of hardening at continuous fluences than the TM 316L SS. At a dose of 6 dpa, the hardening rates of the SLM and TM 316L SS specimens were 102.1 % and 84.1 %, respectively. Furthermore, TEM results showed that the density of dislocation loops in the TM 316L SS was 8.36 × 1023/m3 while that in the SLM 316L SS was 6.15 × 1023/m3. The average sizes of the dislocation loops were 1.35 and 1.22 nm, respectively. According to calculations using the dispersed barrier hardening (DBH) model, the dislocation loops were the primary factor influencing the irradiation hardening behavior. Under Fe-ion irradiation at RT, SLM 316L SS showed superior resistance to irradiation hardening and swelling compared with TM 316L SS.

Revealing the temperature-dependent hardening mechanisms of non-uniformly distributed defects for ion-irradiated metals: Experiments and modeling

In this work, an integrated approach including experimental examinations and theoretical modeling was employed to examine the temperature-dependent hardening of structural materials under ion irradiation. Experimental procedures involved the characterization of grain morphology and size distribution by electron backscatter diffraction (EBSD). Irradiation experiments were conducted utilizing high-energy Au-ions at elevated temperatures, with transmission electron microscopy (TEM) employed for observing irradiation-induced defects. Moreover, mechanical properties were assessed through high-temperature nano-indentation tests (HTNIT), revealing the concurrent occurrence of irradiation hardening and high-temperature softening in the considered materials. For theoretical modeling, a mechanistic model was developed to address the depth-dependent hardness. This model innovatively characterized the hardening mechanisms resulting from non-uniformly distributed defects in the irradiated layer while simultaneously incorporating the temperature effect. Calibration of the model demonstrated its capability to elucidate the hardness-depth relationships at different irradiation doses and testing temperatures. Further theoretical analyses indicated distinct evolution laws in irradiation hardening behavior with varying indentation depths, while the increment in temperature accelerated the expansion of the plastic zone, resulting in a moderate indentation size effect (ISE). With increasing indentation depth, the hardening contributions of geometrically necessary dislocations (GNDs) and irradiation-induced defects gradually diminished. At substantial indentation depths, the dominant hardening mechanism was then controlled by statistically stored dislocations (SSDs).

The influence of solute elements on grain size, hardness and irradiation-induced defect clusters in dilute Fe-based alloys

The precipitation of solutes is the primary factor that leads to the occurrence of irradiation hardening and embrittlement, thus limiting the service life of reactor pressure vessel (RPV) steels. In this study, four model alloys (Fe-1.35Mn-0.75Ni, Fe-1.35Mn-0.20Si, Fe-0.75Ni-0.20Si and Fe-1.35Mn-0.75Ni-0.20Si) were used to investigate the influence of solute elements on grain size, hardness, irradiation hardening, and defect clustering behavior. The materials were irradiated with 3.5 MeV Fe ions to doses of 0.1, 0.3, 1.0, and 3.0 dpa at 290 °C. Results indicated that the addition of solute Mn resulted in grain refinement while the addition of solutes Ni and Si led to grain coarsening. The strength of the model alloys was affected by the change in grain size. A definite dose rate effect was found in all four model alloys under irradiation, where irradiation with a higher dose rate led to a lower irradiation hardening effect and irradiation with a lower dose rate induced a higher hardening effect. Atom probe tomography (APT) results revealed that the precipitated solutes were always coupled with interstitial defect clusters. Solute Mn had a strong influence on cluster nucleation and a weak influence on cluster size. Solutes Si and Ni significantly promoted the nucleation and growth of clusters, and Ni had a greater influence on the nucleation and growth of clusters than Si. Meanwhile, solute Si consistently exhibited a low fraction in the clusters when the solute Mn element was presented in the alloys.

Ion-irradiation induced hardening behavior of zirconium alloys: A combination of experimental and theoretical study

This study aimed to investigate the microstructure information and mechanical behavior of ion-irradiated zirconium (Zr) alloys through a combination of experimental measurements and theoretical analysis. Experimental procedures involved Au3+ irradiation of Zr-1.5Sn-0.2Fe-0.1Cr and Zr-1.0Sn-1.0Nb-0.1Fe at 3 displacements per atom (dpa) and 30 dpa. Transmission electron microscopy analysis revealed the formation of precipitates such as Zr(Fe, Cr)2 in Zr-1.5Sn-0.2Fe-0.1Cr, as well as β-Nb and Zr(Nb, Fe)2 in Zr-1.0Sn-1.0Nb-0.1Fe. These precipitates exhibited a transformation from crystalline state to partial or complete amorphization after ion-irradiation. Furthermore, macroscopic mechanical properties of the Zr alloys were characterized by nano-indentation tests, which revealed a significant indentation size effect (ISE) and irradiation hardening behavior. To provide a theoretical understanding of the depth-dependent hardness observed in ion-irradiated Zr alloys, a mechanistic model was developed to address the different expansion rates of the plastic zone in the irradiated and unirradiated region. The results indicated that the ISE phenomenon could be attributed to the reduction in the density of geometrically necessary dislocations (GNDs) caused by the expansion of the plastic zone. In addition, the irradiation hardening behavior primarily resulted from the contribution of irradiation-induced defects. With increasing indentation depth, the dominant hardening mechanisms changed from the contribution of irradiation-induced defects, GNDs and statistically stored dislocations (SSDs) in the irradiated region to SSDs in the unirradiated substrate.

Irradiation hardening and void swelling behaviors of F82H IEA by using dual-ion irradiation

ABSTRACT The irradiation hardening behavior and the void swelling behavior of the Japanese reduced-activation ferritic/martensitic steel, i.e. F82H IEA heat, were investigated by using single- and dual-ion irradiation experiments in this study. For irradiation hardening of F82H, a map of the hardness ratio before/after ion irradiation was obtained for 270–500 °C up to 80 dpa. The peak temperature of irradiation hardening was at about 350°C. On the other hand, it was shown that an extra hardening occurred in helium co-implanted F82H for 400–430 °C except for the temperatures less than 400 °C and more than 450 °C. For the temperature dependence of void swelling in helium co-implanted F82H, it was confirmed that the peak swelling temperature was identified at about 470 °C. That amount of swelling (average value) was 0.4% at 20 dpa.

Irradiation hardening behavior of high entropy alloys using random field theory informed discrete dislocation dynamics simulation

High entropy alloys (HEAs) exhibit outstanding radiation resistance over traditional alloys owing to severe lattice distortion, and therefore, would be regarded as candidate materials for advanced reactors. However, the complex evolution of dislocation configurations in the irradiated HEAs subjected to the external loads has not been uncovered at the micron scale. To elucidate this key mechanism, we develop the random field theory informed discrete dislocation dynamics simulations based on high-resolution transmission electron microscopy, to systematically clarify the role of heterogeneous lattice strain on the complex interactions between the dislocation loop and dislocation in three-dimensional (3D) space. The results show that the lattice-strain-induced irradiation hardening decreases, in agreement with the excellent irradiation hardening resistance of the HEAs from the recent experiment. Based on the analysis of the micrometer-scale dislocation multiplication process, a new cross-slip mechanism through the collinear reaction between dislocations and rhombus perfect loops is revealed. The peaks and valleys of the heterogeneous lattice strains induced by severe lattice distortion are randomly distributed in a 3D space, reducing an aggregation of lattice point and dislocation density at the local region. The cross-slip process occurs by the joint action of the lattice distortion and rhombus-perfect-loop dislocation reaction, leading to the strong irradiation resistance attributed to a large number of narrow defect-free channel under a relatively random strain localization. The present study gives an insight into the fundamental irradiation damage behavior at the mesoscopic scale, thereby guiding the development of advanced HEA materials through the regulation of heterogeneous lattice strain for nuclear energy applications.

Temperature-dependent hardening of zirconium alloys under heavy ion-irradiation: Experimental observation and theoretical analysis

In this work, microstructure observation and thermo-mechanical property characterization of ion-irradiated Zr-0.5Sn-0.5Nb-0.3Fe-0.015Si alloys were addressed by both experimental measurements and theoretical analyses. In experiments, 6 MeV Au3+ irradiation was carried out for Zr alloys with the irradiation dose up to 30 displacements per atoms (dpa). Transmission electron microscopy (TEM) indicated the addition of solute elements could facilitate the formation of second phase particles (SPPs) like (Zr, Nb)2Fe, Zr(Nb, Fe)2 and Zr-Nb-Fe-Si precipitates in the unirradiated matrix, which were then transformed from the crystalline to amorphous state during the irradiation process. Moreover, it was indicated that there existed microband structures with different orientations inside the grains, and the actual irradiation depth had exceeded the value predicted by the SRIM simulation. The influence of irradiation-induced defects on the macroscopic mechanical behaviors of Zr alloys was characterized by both nano-indentation test (NIT) at 298 K and high-temperature nano-indentation test (HTNIT) at 573 K and 673 K. To theoretically address the hardness-indentation depth relations of ion-irradiated Zr alloys at elevated temperatures, a mechanistic model considering the underlying hardening mechanisms was developed. It was revealed that the irradiation hardening behavior was attributed to the contribution of irradiation-induced defects, while the decrease of hardness at elevated temperatures was ascribed to the weakening of all hardening mechanisms. In addition, with increasing indentation depth, the dominant hardening mechanisms changed from the contribution of irradiation-induced defects and geometrically necessary dislocations (GNDs) to those of statistically stored dislocations (SSDs).

Irradiation hardening of stainless steel model alloy after Fe-ion irradiation and post-irradiation annealing treatment

Microstructural observation with atom probe tomography, transmission electron microscopy, and mechanical examination with a nano-indenter were performed on stainless steel model alloy that had been irradiated with Fe ions and subjected to post-irradiation annealing treatment (PIA). Significant irradiation hardening occurred in the as-irradiated specimen and irradiation hardening recovery occurred at 400 °C of cumulative PIA, Tiny black dots, Frank loops and Ni-Si clusters formed in the as-irradiated specimen and their recovery was observed with the cumulative PIA. A classic dispersed-barrier hardening model was used to evaluate irradiation hardening connected to microstructural information by applying the density and size of defect variants. Ni-Si clusters and Frank loops contributed substantially to the increase of irradiation hardening, whereas the black dots contributed little for the stainless steel model alloy. The barrier strength factor of Ni-Si clusters, αSC, was obtained as 0.029 and 0.015 for two different obstacle strength model, which suggests that an individual Ni-Si cluster is a weak barrier obstacle to mobile dislocation on the dispersed-barrier hardening model, but that the dense formation of Ni-Si clusters controls irradiation hardening behavior. A quantitative analysis was performed of the increase in strength from the Ni-Si clusters that was caused by coherency strains, and an irradiation hardening mechanism for the stainless steel, including the solute atom clusters, is discussed.

Correlation between Microstructural Change and Irradiation Hardening Behavior of He-Irradiated V–Cr–Ti Alloys with Low Ti Addition

V–4Cr–xTi (x = 0 to 4) alloys were used to investigate the additional effect of Cr, Ti and interstitial impurities on the microstructural evolution in He-irradiated V–Cr–Ti alloys to minimize radioactivity after fusion neutron irradiation. Transmission electron microscopy and atom probe tomography were carried out to the He-irradiated specimens at 500 °C with 0.5 dpa at peak damage. A flash electro-polishing method for the FIB-extracted specimen was established for the ion-irradiated vanadium alloys. The microstructural evolution of the irradiation-induced titanium-oxycarbonitride, Ti(CON) precipitates was observed and was influenced by the effect of Ti addition on the Ti(CON) precipitation. Apparent Ti(CON) precipitates formed in V-4Cr-xTi with 2% addition of Ti. In the V-4Cr-1Ti alloy, a high density Ti enriched cluster was formed. The origin of the irradiation hardening increase resulted from the size distribution of Ti(CON) precipitation from the dispersed barrier-hardening theory.

Irradiation Hardening Behavior of He-Irradiated V–Cr–Ti Alloys with Low Ti Addition

A set of V–(4–8)Cr–(0–4)Ti alloys was fabricated to survey an optimum composition to reduce the radioactivity of V–Cr–Ti alloys. These alloys were subjected to nano-indenter tests before and after 2-MeV He-ion irradiation at 500 °C and 700 °C with 0.5 dpa at peak damage to investigate the effect of Cr and Ti addition and gas impurities for irradiation hardening behavior in V–Cr–Ti alloys. Cr and Ti addition to V–Cr–Ti alloys for solid–solution hardening remains small in the unirradiated V–(4–8)Cr–(0–4)Ti alloys. Irradiation hardening occurred for all V–Cr–Ti alloys. The V–4Cr–1Ti alloy shows the highest irradiation hardening among all V–Cr–Ti alloys and the gas impurity was enhanced to increase the irradiation hardening. These results may arise from the formation of Ti(CON) precipitate that was produced by He-ion irradiation. Irradiation hardening of V–Cr–1Ti did not depend significantly on Cr addition. Consequently, for irradiation hardening and void-swelling suppression, the optimum composition of V–Cr–Ti alloys for structural materials of fusion reactor engineering is proposed to be a highly purified V–(6–8)Cr–2Ti alloy.

Simulation of Mechanical Property and Fracture Behavior of Reactor Pressure Vessel Steel by Small Punch Test

The life extension of the reactor requires the improvement of the material performance, especially the radiation resistance. At present, some data about the irradiation hardening and embrittlement behavior of reactor pressure vessel steel have been obtained, mainly the standard tensile properties, standard impact properties and so on. However, to understand the service behavior of materials at higher doses, higher neutron fluence radiation tests are needed. The problem is that because of the large size of the standard sample, if the irradiation dose continues to increase, the surface dose of the irradiated sample will be larger, and it is difficult to test after irradiation. In addition, if the irradiation sample can be replaced by smaller sample, the utilization space of the radiation channel will be greatly improved. The high-throughput characterization for irradiation performance of reactor pressure vessel steel can be achieved. Therefore, in this work, the small punch test of reactor pressure vessel material A508-3 steel was carried out at the temperatures ranging from -150 °C to room temperature. The performance data, such as tensile characteristics, yield strength and fracture energy, were obtained from load-deformation curves. Comparing the results of small punch test with the data obtained by standard tensile and impact test, the correlation on the performance data was established for A508-3 steel.

Vickers hardness change of the Chinese low-activation ferritic/martensitic steel CLF-1 irradiated with high-energy heavy ions

In the present work, the irradiation hardening behavior of a Chinese low-activation ferritic/martensitic steel CLF-1, a candidate for fusion reactor blankets, is studied. Specimens were irradiated with high-energy 14N and 56Fe ions at the terminal of a cyclotron to three successively increasing damage levels of 0.05, 0.1 and 0.2 displacements per atom (dpa) at about −50 °C. The energy of the incident ions was dispersed to 11 successively decreasing grades using an energy degrader, thereby generating an atomic displacement damage plateau in the specimens from the surface to a depth of 25 μm, which is sufficiently broad for the Vickers hardness test. Eight different loads (i.e. 98 mN, 196 mN, 490 mN, 980 mN, 1.96 N, 4.9 N, 9.8 N and 19.6 N) were applied to the specimens to obtain the depth profiles of the Vickers hardness by using a micro-hardness tester. Hardening was observable at the lowest damage level, and increased with increasing irradiation dose. A power-law correlation of the Vickers hardness with the damage level (HV0 = 1.49 + 0.76 dpa0.31) is proposed. Testing with a nano-indentation technique was also performed, and a linear relationship between the Vickers micro-hardness and the nano-hardness (HV0 = 0.83H0) was observed. A comparison with other RAFM steels (CLAM, JLF-1, F82H, EUROFER97 etc.) under neutron or charged particle irradiation conditions shows that most of the RAFM steels exhibit similar power-law exponents in the dose dependence of irradiation hardening. The difference in the irradiation hardening may be attributed to differences in microstructure prior to irradiation.

Irradiation hardening induced by blistering in tungsten due to low-energy high flux hydrogen plasma exposure

In this work, the microstructure evolution in the near-surface of tungsten under hydrogen (H) plasma exposure conditions was observed by means of scanning electron microscopy (SEM), plasma focused ion beam (FIB) and transmission electron microscopy (TEM) techniques. Blisters, with existing dislocations distributed around obviously, were observed beneath the tungsten surface when the exposure temperature was 573 K, which was rarely reported in previous studies. However, H bombardment at 1273 K did not lead to the formation of blister-like microstructures. Correspondingly, irradiation hardening occurred after low temperature exposure, but not after high temperature exposure, according to the Berkovich nano-indentation experiments. In order to characterize the indentation size effect and irradiation hardening behavior of plasma-exposed materials, a mechanistic model was proposed for the hardness-depth relationship. A good agreement between the experimental indentation data and theoretical results revealed that plasma-induced dislocations play a dominant role in determining the increase of hardness for H plasma-exposed tungsten.

Irradiation effects in high entropy alloys and 316H stainless steel at 300 °C

High entropy alloys (HEAs) have been considered for applications in nuclear reactors due to their promising mechanical properties, corrosion and radiation resistance. It has been suggested that sluggish diffusion kinetics and lattice distortion of HEAs can enhance the annihilation of irradiation-induced defects, giving rise to a higher degree of tolerance to irradiation damage. In order to understand the irradiation effects in HEAs and to demonstrate their potential advantages over conventional austenitic stainless steels (SS), we performed in-situ ion irradiation experiments with 1 MeV krypton at 300 °C on two HEAs and a 316H SS under an identical irradiation condition. The irradiation introduced a high density of dislocation loops in all materials, and the microstructural evolution as a function of dose was similar for HEAs and 316H SS. Nanoindentation tests showed that the degree of irradiation hardening was also comparable between them. The similar microstructural evolution and irradiation hardening behavior between the HEAs and 316H indicate that, at low temperatures (≤300 °C), the irradiation damage of fcc alloys is not sensitive to compositional variation and configurational entropy.

A hierarchical theoretical model for mechanical properties of lath martensitic steels

Lath martensitic steels possess excellent mechanical properties because of the special microstructure, i.e. laths, blocks and packets with particular crystallography in a prior austenite grain. To theoretically analyze the mechanical behavior of lath martensitic steels, a hierarchical model is proposed combining the crystal plasticity theory and micromechanical method. Within the crystal plasticity theory, an interfacial dislocation model is proposed at the block level to physically describe the effect of lath boundary on the deformation behavior. Furthermore, The scale transition among block, prior austenite grain and lath martensitic steel is accomplished by the elastic-viscoplastic self consistent theory. Based on the proposed model, the deformation behavior of lath martensitic steel with hierarchical structures has been theoretically analyzed. It is revealed that the deformation anisotropy of prior austenite grain is weak due to the particular crystallography of microstructures, and the influence of temperature on yield stress is dominated by the thermally related intrinsic lattice friction and elastic modulus. Moreover, the hierarchical model is generalized to study the mechanical behavior of irradiated lath martensitic steel, and it is found that the change of temperature has a limited effect on the irradiation hardening behavior, and the increase of yield stress induced by irradiation-induced defects can accelerate the evolution rate of dislocation density. Numerical results with/without irradiation effect can both match well with corresponding experimental data, indicating the good accuracy and rationality of the proposed model.

Effect of neutron irradiation on rhenium cluster formation in tungsten and tungsten-rhenium alloys

To clarify the irradiation hardening behavior of neutron-irradiated tungsten(W), the formation behavior of rhenium(Re)-enriched cluster and W-Re precipitation with displacement damage in fission reactors environment was investigated by the atom probe tomography (APT) technique, and the results obtained by APT and microstructural observation by transmission electron microscopy (TEM) were compared. Neutron-irradiated pure W and W-10Re alloys were used for the analysis. These samples were irradiated to 0.96 dpa at 538 °C in a fast experimental reactor Joyo, and irradiated to 0.90 dpa at 500 °C in a mixed spectrum reactor HFIR. The chemical compositions of the matrix, Re clusters, and precipitates were determined by the AP analysis. The Joyo-irradiated samples contained lower levels of transmuted Re and Os, while the HFIR-irradiated samples contained higher levels of Re and Os. The Re-enriched clusters demonstrated characteristic spatial distribution related to defect cluster distribution in each irradiated specimen.

The effect of He bubbles on the swelling and hardening of UNS N10003 alloy

In order to investigate the effect of He bubbles on the swelling and hardening of nickel based alloys, 1.2 MeV He ion irradiation was performed on UNS N10003 alloy at 650 °C. The number densities and sizes of He bubbles increase with increasing irradiation dose. Neither “black spots” nor dislocation loops were observed in TEM images of the as-irradiated samples. Irradiation dose up to 6.18 displacements per atom (dpa) causes a swelling of 2.67%, which is mainly ascribed to He bubbles. He bubbles induced hardening increases with increasing irradiation dose. The irradiation hardening behavior was compared with heavy ion irradiation. The result reveals that the obstacle strength of He bubbles is less than that of the saturated heavy ion induced irradiation defects.

Investigation of hardening behavior in Xe ion-irradiated Zr–1Nb

Irradiation hardening behavior of Zr–1Nb was examined by nanoindentation, slow positron annihilation technique, transmission electron microscopy and coplanar extremely asymmetric X-ray diffraction technique. Samples were irradiated at a dose rate of 2.78 × 10−4 dpa/s to peak doses of 0.15, 0.5, 1.5 and 2.5 dpa with 6.37 MeV Xe26+ ion beam at room temperature. The increase in hardness as a function of dose followed a power law expression with the exponent of 0.46. With increasing irradiation dose, more mono-, di- and trivacancies were induced, but their concentration remained constant once formed due to the equilibrium between the formation and recombination of vacancy type clusters during irradiation. Meanwhile, the dislocation loops were also introduced and their linear density increased with dose. The dislocation loops played an important role in the irradiation hardening behavior. But the exact contribution of each microstructural components to the overall hardness still needs further study.

Helium and hydrogen irradiation induced hardening in CLAM steel

The irradiation hardening behavior of low activation martensitic steel was investigated by nano-indentation measurements combined with transmission electron microscopy analysis. The steel was irradiated by single-(He+) and sequential-(He+ plus H+ subsequently) ions at 773K respectively. The continuous hardness (H) vs. depth (h) profiles at shallow depths were given by the continuous stiffness measurement (CSM) which can be analyzed according to Nix–Gao model and film/substrate system hardness model. The nano-indentation hardness results showed that both single-(He+) and sequential-(He+ plus H+ subsequently) irradiated steel exhibited obvious hardening and the hardening level of single-(He+) was higher than that of sequential-(He+ plus H+ subsequently) one. The bulk-equivalent hardness H0irr of the ion irradiated region was obtained by fitting the nano-indentation data using Kasada method. The microstructure analysis indicates that sequential-ion irradiation induced dislocation loops with larger size and smaller number density compared to single-ion irradiation at 773K. The irradiation induced hardening behavior in the steels subjected to different ion irradiations can be explained by the dispersed barrier hardening model which was consistent with the nano-indentation hardness results.

Depth-dependent nanoindentation hardness of reduced-activation ferritic steels after MeV Fe-ion irradiation

The irradiation hardening behavior of F82H reduced-activation ferritic steels after MeV Fe-ion beam irradiation experiments was investigated with a nanoindentation test. Two sets of ion-irradiation experiments were conducted at 270°C with 10.5MeV Fe3+ ions up to 5dpa at a 1000nm depth at TIARA facility and at 290°C with 6.4MeV Fe3+ ions up to 3dpa at a 600nm depth at DuET facility, respectively. The measured nanoindentation hardness was converted to the depth-dependent bulk-equivalent hardness based on a combination of the Nix–Gao model to explain the indentation size effect and the film/substrate model to explain the damage gradient effect in the ion-irradiated region and softer substrate effect of the non-irradiated region beyond the irradiated depth range.