Irradiation-induced Hardening Research Articles

Orientation-dependent ion-irradiation responses in molybdenum and molybdenum-rhenium alloys

Helium ion-irradiation responses in molybdenum (Mo) and molybdenum-rhenium (Mo-Re) alloys at room temperature were evaluated via micro-pillar compression. Irradiation-induced hardening is closely dependent on orientation in Mo-Re alloys with low Re content, while is weak orientation-dependence on high Re alloys. Irradiated pillars with [100] orientation all shows higher flow stresses than pillars with [110] orientation after yield, which is distinguishing from pillars before irradiation that has been shown in our previous work. High density of dislocation loops and helium bubbles after irradiation at room temperature induces severe hardening in Mo. Voids / small loops induced by irradiation lead to less hardening in Mo-Re alloy than in Mo. No precipitation is observed in all samples, which may be attributed to low irradiated temperature and implantation energy.

A multiscale study on the microstructure and hardening models of the irradiation defects on reactor pressure vessel steels: Modelling and experiment

Irradiation defects are known to impede the dislocation motion in irradiated reactor pressure vessel (RPV) steels. However, consistent experimental studies of the respective irradiation defects and relevant theoretical studies of the interaction mechanism and hardening model are still lacking. In this study, the microstructure and irradiation-induced hardening of RPV steels are investigated using transmission electron microscopy (TEM)/energy-dispersive X-ray spectrometry (EDS) and molecular dynamics (MD) simulations. In A508-Ⅲ steel, 9R, twinned fcc, and other typical Cu-rich cluster structures are observed using TEM after the irradiation experiment. Based on the experimental results, MD simulations are performed to study the interactions between the dislocations and Cu clusters or voids. A mechanism diagram of the interaction mechanism coupling the effects of temperature and defect size is shown for both the Cu clusters and voids. A relevant hardening model is proposed based on the MD data, which is further discussed in relation to the macroscopic mechanical properties. Furthermore, the irradiation-induced hardening of the RPV steel is successfully predicted, which further demonstrates the reliability of the proposed unified hardening model.

Ultrafast annihilation of irradiation-induced defects using pulsed electric current for damage performance regeneration

As the most important irradiation-induced defects, dislocation loop and copper-rich nanocluster are the major contributors to the embrittlement of the neutron-irradiated reactor pressure vessel steels. In this study, such nano-defects were introduced into the material by 3 MeV Fe ions up to the dose of 1 dpa at high temperature (290 °C) to simulate neutron irradiation. It was found that pulsed electric current can effectively reduce 95 % of irradiation-induced hardening. Correspondingly, the characterization results showed that almost all the dislocation loops disappeared and the quantity of copper-rich nanoclusters also reduced greatly at relatively low temperature (450 °C), and the process took only 20 min. Meanwhile, it was qualitatively proved by positron annihilation spectroscopy that the number of irradiation-induced vacancy-type defects and solute-enriched clusters was significantly decreased after electropulsing. Furthermore, under the pulsed electric field, the rapid annihilation of the dislocation loops due to their accelerated collision with vacancies can remove the nucleation sites of the copper-rich nanoclusters and make them become dispersed, further promoting the nanoclusters that lack nucleation sites dissolving faster. Therefore, this electropulsing treatment provides a practical ‘‘in-situ” performance repair technology to extend the service life of reactor pressure vessel steels by regulating the interaction between vacancies, interstitial atoms and irradiation-induced defects.

Microstructure-Informed Prediction of Hardening in Ion-Irradiated Reactor Pressure Vessel Steels

Ion irradiation combined with nanoindentation is a promising tool for studying irradiation-induced hardening of nuclear materials, including reactor pressure vessel (RPV) steels. For RPV steels, the major sources of hardening are nm-sized irradiation-induced dislocation loops and solute atom clusters, both representing barriers for dislocation glide. The dispersed barrier hardening (DBH) model provides a link between the irradiation-induced nanofeatures and hardening. However, a number of details of the DBH model still require consideration. These include the role of the unirradiated microstructure, the proper treatment of the indentation size effect (ISE), and the appropriate superposition rule of individual hardening contributions. In the present study, two well-characterized RPV steels, each ion-irradiated up to two different levels of displacement damage, were investigated. Dislocation loops and solute atom clusters were characterized by transmission electron microscopy and atom probe tomography, respectively. Nanoindentation with a Berkovich indenter was used to measure indentation hardness as a function of the contact depth. In the present paper, the measured hardening profiles are compared with predictions based on different DBH models. Conclusions about the appropriate superposition rule and the consideration of the ISE (in terms of geometrically necessary dislocations) are drawn.

Assessing the impact of irradiation-induced defects on the hardening of reactor pressure vessel materials for sustainable development of nuclear energy

The operational longevity of a nuclear power plant, specifically of the light water reactor type, hinges on the assessment of component degradation, with a particular focus on the reactor pressure vessel (RPV) materials. Among the most prevalent defects responsible for irradiation-induced hardening in RPV materials are point defects (PDs) and dislocation loops (DLs). However, their respective contributions to the overall hardening of RPV materials have not been clearly determined. To address this issue, this study was conducted using iron alloys with equivalent Mn and Ni content to those used in nuclear RPV in Japan. The study employed transmission electron microscopy (TEM), atom probe tomography (APT), and nanoindentation to assess the roles of PDs and DLs in contributing to the hardening behaviour of RPV materials. The results indicated that both solute clusters (SCs), which are formed by the aggregation of PDs, and DLs play equally important roles in the hardening behaviour of RPV materials. However, an in-depth analysis of three-dimensional images using the APT method revealed that there may have been errors in evaluating SCs and DLs, leading to double counting when assessing the impact of irradiation-induced defects on the hardening behaviour of RPV materials. These findings highlight the need for improved accuracy in evaluating these defects to better understand the behaviour of RPV materials under irradiation.

He bubbles inhibition and abnormal hardening in GH3535 alloy by niobium element addition

GH3535 and 2 wt.% Nb-modified GH3535 (Nb-GH3535) alloys were irradiated with 1.2 MeV He ion at fluences of 1 × 1016, 3 × 1016 and 1 × 1017 ions/cm2 at 750 °C. Microstructural changes, including the formation of He bubbles and superlattice phases, as well as the irradiation-induced hardening, were investigated by TEM and nanoindentation. Notably, Nb-GH3535 exhibited a smaller size and lower number density of He bubbles, especially at higher fluences, indicating the inhibitory effect of Nb on bubble evolution. First-principles calculations revealed that the presence of Nb inhibits the combination of He atoms and vacancies, reducing their diffusion rate and consequently impeding the nucleation and growth of He bubbles. Furthermore, the addition of Nb facilitates the irradiation-induced precipitation of the superlattice phase, leading to a higher degree of total hardening in Nb-GH3535 alloy.

Irradiation-induced hardening of Zircaloy-2 at room temperature under external stress conditions

Irradiation-induced hardening of Zircaloy-2 at room temperature under external stress conditions

HARDENING OF LIGHTWEIGHT MULTI-PRINCIPAL ELEMENT TITANIUM-BASED ALLOY UNDER AR ION IRRADIATION

Among new prospective materials multi-principal element alloys (MPEA) have attracted considerable attention in recent years due to their excellent corrosion and irradiation resistance as well as their good mechanical properties over a wide temperature range. The new lightweight multi-principal element titanium-based alloy 61Ti-10Cr-7Al11V-11Nb (at. %) with high ductility at room and elevated temperatures is studied. This single-phase bcc alloy was irradiated with 1.4 MeV Ar ions at room temperature and mid-range doses from 1 to 10 displacements per atom. The effect of irradiation is studied by examining the hardening. A comparison was performed with irradiation-induced hardening behaviour of MPEA, 316 austenitic stainless steel irradiated under an identical condition. It was shown that hardness increases with irradiation dose for all the materials studied, but this increase is lower in multi-principal element alloys both face-centered cubic (FCC) and body-centered cubic (BCC) structures than in stainless conventional steel.

The design of Pd-containing high-entropy alloys and their hardening behavior under He ion irradiation

Imparting additional chemical heterogeneities of high-entropy alloys (HEAs) is considered as a novel strategy to improve their irradiation resistance on dislocation loop growth; however, there is still a lack of knowledge about the effects of this strategy on bubble evolution and irradiation-induced hardening. In this study, a series of Pd-containing FeCrNiCo HEAs with proven potential for chemical short-range order were designed and irradiated by 400 keV He ions to the fluences of 5 × 1016 cm−2, 1 × 1017 cm−2, and 5 × 1017 cm−2 at 723 K. Microscopic characterization demonstrates a remarkably inhibited growth of dislocation loops and promoted bubble coarsening in the Pd-containing HEAs compared to the HEA without Pd alloying. The underlying mechanism is that although the alloying of Pd can suppress the evolution of dislocation loops through severe local lattice distortion, it also significantly reduces the vacancy formation energy of the entire HEA system, leading to the encouraged bubble evolution. Pd-containing HEAs exhibit improved resistance to irradiation-induced hardening at low doses, but this advantage is unexpectedly absent with increasing dose. The hardening mechanism is explained by identifying the loop-dominated hardening at low dose and the bubble-induced hardening which makes an increased contribution at high dose. In addition, the Ni segregation around bubbles and loop punching mechanism are also discussed to provide essential evidence. These insights can help to shed light on the effects of Pd alloying on bubble evolution and irradiation-induced hardening behavior in HEAs, and thus contribute to the design of irradiation-resistant HEAs.

Displacement cascade under high-energy irradiation and tensile property in Fe-Cr-Al alloys: A molecular dynamics simulation

Fe-Cr-Al alloy is considered as one of the most prospective accident-tolerant fuel cladding materials due to its excellently comprehensive properties. Irradiation performance evaluation of Fe-Cr-Al alloy is the key for successful development of fuel-clad system with enhanced accident tolerance. However, the microstructural response is largely dependent on irradiated conditions. In the present work, molecular dynamic method is employed to simulate the cascade collision processes under high-energy irradiation and tensile deformation behaviors of Fe-Cr-Al alloys. The threshold displacement energies (TDEs) in Fe-Cr-Al alloys are calculated at different temperatures and along various directions, and the TDEs along different directions exhibit significant anisotropy. The effect of alloy compositions on the displacement cascade of Fe-Cr-Al alloys is systematically studied, as well as the direction of the primary knock-on atom (PKA), temperature, and PKA energy. At 300 K and 700 K, no significant dependence between the number of surviving Frenkel pairs and PKA energies is observed in Fe-Cr-Al alloys under high-energy irradiation. High temperature and high-energy PKA could potentially lead to form large-sized interstitial clusters and a few small-sized vacancy clusters in Fe-Cr-Al alloys. Dislocations or dislocation loops can easily form at high temperatures and in high-energy irradiation. The effect of alloy compositions, temperatures and PKA energies on the tensile properties of Fe-Cr-Al alloys before and after irradiation, is systematically evaluated. By analyzing the corresponding yield strength, ultimate tensile strength and elastic modulus, Fe-Cr-Al alloys exhibit the characteristic of irradiation-induced hardening and embrittlement with the increase of PKA energy. As the temperature increases, the rigidity of alloys decreases while the irradiation embrittlement significantly increases.

Modelling FeCrAl cladding thermo-mechanical performance. Part II: Comparative analysis with Zircaloy under LOCA conditions

The demonstration of the enhanced thermo-mechanical response of iron-chromium-aluminum (FeCrAl) cladding under any circumstances, particularly under postulated accident scenarios, requires the development of analytical capabilities. By extending FRAPTRAN-2.0 to model the instantaneous plasticity and high-temperature creep of FeCrAl, among other properties, its performance under LOCA conditions has been assessed in comparison with that of Zircaloy-4. Based on the conditions set in the IFA-650.10 test of the HALDEN project, FeCrAl has shown an enhanced thermo-mechanical response due mainly to its superior mechanical strength, which led to a significant extension of the time to failure, even with a reduced cladding thickness. This time gain is expected to be even longer given that irradiation-induced hardening (not considered in the present model) might further delay the onset of plastic deformation. This effect is planned to be implemented in the future in-house extension of FRAPTRAN-2.0.

The Influence of Coherent Oxide Interfaces on the Behaviors of Helium (He) Ion Irradiated ODS W.

Tungsten (W), as a promising plasma-facing material for fusion nuclear reactors, exhibits ductility reduction. Introducing high-density coherent nano-dispersoids into the W matrix is a highly efficient strategy to break the tradeoff of the strength-ductility performance. In this work, we performed helium (He) ion irradiation on coherent oxide-dispersoids strengthened (ODS) W to investigate the effect of coherent nanoparticle interfaces on the behavior of He bubbles. The results show that the diameter and density of He bubbles in ODS W are close to that in W at low dose of He ion irradiation. The radiation-induced hardening increment of ODS W, being 25% lower than that of pure W, suggests the involvement of the coherent interface in weakening He ion irradiation-induced hardening and emphasizes the potential of coherent nano-dispersoids in enhancing the radiation resistance of W-based materials.

Improved irradiation resistance of Cr-Fe alloy for Cr-coated Zircaloy application in accident tolerant fuel

Cr coating onto Zircaloy for the application of accident tolerance fuel claddings became a popular topic in the nuclear material field since the Fukushima-Daiichi accident in 2011. However, in the Cr coating layer the solute element effect on the potential performance improvement is still unknown, especially under irradiation. The irradiation effect in Cr-7Fe was studied after irradiation by 2.8 MeV Fe2+ to 20 dpa at 550 °C. The nano-indentation results show that doping of Fe not only can strengthen the material but also can significantly reduce the irradiation-induced hardness compared with pure Cr. The microstructure analyses show that solute element Fe can retain the diffusion of defects and reduce void swelling. In addition, the APT results indicate that no detrimental Fe-rich cluster formed in irradiated Cr-7Fe. From the aspects of irradiation-induced hardening, void-swelling, and clustering, a great improvement in irradiation resistance was achieved in Cr-7Fe alloy compared with pure Cr.

Combined effects of irradiation and hydrogen on the mechanical and corrosion performances of the ferrite in duplex phase steels

Structural materials in nuclear reactor cores suffer from severe corrosion and weakened mechanical properties under the combined influences of hydrogen and irradiation. In this study, we studied the peak damaged region (PDR) in the ferrite phase of a duplex stainless steel post proton irradiation, which is featured with a relatively higher concentration of hydrogen and level of irradiation damages than that of the uniformly damaged region (UDR). It is found that the irradiation-induced hardening effect is more evident in UDR than in PDR, and the PDR exhibits a much larger drop of electron work function and worse corrosion resistance than the UDR. Moreover, the oxide film on the PDR shows a much lower electrical resistivity than that of the UDR. The degradation of mechanical performance and corrosion resistance is associated with the local enrichment of hydrogen and the formation of G-phase in the PDR under the unique combined influence of hydrogen and irradiation. This study sheds lights on the synergetic effects of hydrogen and irradiation on the degradation of steels in reactor cores.

A Novel BCC-Structure Zr-Nb-Ti Medium-Entropy Alloys (MEAs) with Excellent Structure and Irradiation Resistance.

Medium-entropy alloys (MEAs) are prospective structural materials for emerging advanced nuclear systems because of their outstanding mechanical properties and irradiation resistance. In this study, the microstructure and mechanical properties of three new single-phase body-centered cubic (BCC) structured MEAs (Zr40Nb35Ti25, Zr50Nb35Ti15, and Zr60Nb35Ti5) before and after irradiation were investigated. It is shown that the yield strength and elongation after fracture at room temperature are greater than 900 MPa and 10%, respectively. Three MEAs were irradiated with 3 MeV Fe11+ ions to 8 × 1015 and 2.5 × 1016 ions/cm2 at temperatures of 300 and 500 °C, to investigate the irradiation-induced hardening and microstructure changes. Compared with most conventional alloys, the three MEAs showed only negligible irradiation hardening and even softening in some cases. After irradiation, they exhibit somewhat surprising lattice constant reduction, and the microstructure contains small dislocation loops. Neither cavities nor precipitates were observed. This indicates that the MEAs have better irradiation resistance than traditional alloys, which can be attributed to the high-entropy and lattice distortion effect of MEAs.

Effect of Au-ion irradiation on the corrosion behavior of CLF-1 steel

The corrosion behavior of the CLF-1 steel irradiated by 6 MeV Au-ions with different fluences was studied by exposing it to a solid tritium breeder (Li4SiO4) in a 0.12 MPa He + 1% H2 atmosphere at 550 °C. The results showed that irradiation-induced hardening appeared in the irradiated CLF-1. The double corrosion layer appeared on the irradiated samples, and the thickness of the corrosion layer increased with the irradiation damage level. A newly formed corrosion product of Li2Fe3O4 appeared on the sample surface corresponding to the highest irradiation damage level. The underlying mechanism for the changed mechanical properties and corrosion products is discussed.

Effectiveness of Thermal Annealing in Recovery of Tensile Properties of Compositionally Tailored PWR Model Steels Irradiated in LYRA-10

Understanding the mechanical behaviour of reactor pressure vessel (RPV) steels at high fluences has become an important topic in regard to Long-Term Operations (LTO) of existing nuclear power plants (NPP). The effectiveness of thermal annealing treatments to recover the mechanical properties of compositionally tailored pressurised water reactor (PWR) model steels irradiated to high neutron fluences, up to 1.22 × 1020 n·cm−2, is analysed in this study. Tensile testing of four different PWR RPV steels was performed after irradiation and subsequent recovery annealing treatment at 450 °C for 40 h. Irradiation-induced hardening and the effectiveness of recovery thermal annealing have been assessed by comparing the strength and ductility properties of irradiated and irradiated and subsequently annealed samples with unirradiated reference samples for all four model steel. The annealing treatment resulted in a significant recovery of the yield strength (~75–89%) and the ultimate tensile strength (~78–96%) of all four PWR model steels. This study proves that substantial irradiation-induced hardening (up to ~389 MPa) observed in steels containing high Ni and Mn contents can still be recovered using the thermal annealing treatment. No influence of annealing on ductility properties has been observed for all four model steels. Microscopy analyses of these steels to understand the underlying irradiation damage and recovery mechanisms are planned for the near future.

Insights into hardening, plastically deformed zone and geometrically necessary dislocations of two ion-irradiated FeCrAl(Zr)-ODS ferritic steels: A combined experimental and simulation study

As one of promising candidate materials for fuel claddings and structural components in the Gen-IV fission reactors, FeCrAl(Zr)-ODS ferritic steels were studied to well understand the radiation hardening behavior. Nanoindentation (NI) hardness and plastically deformed zone (PDZ), geometrically necessary dislocations (GNDs) as well as microstructures were investigated for two FeCrAl(Zr)-ODS ferritic steels irradiated with 6.4 MeV Fe3+ at room temperature (RT) up to the nominal damages of 2, 10 and 50 dpa. Irradiation-induced hardening, which was estimated by using the Nix-Gao model regardless of the damage gradient effect (DGE), increased continuously with increasing the nominal displacement damage. When taking the DGE into account, the dependence of the irradiation hardening (in MPa) on a local damage level (dpa) obtained by finite element method (FEM) simulations was 153.65 × (dpa)0.26 and 158.35 × (dpa)0.25 for the non-Zr steel and the Zr-added one, respectively. The hardening caused by ion-irradiation was discussed in terms of the loss of oxide particles, the formation of dislocation loops and the solid solution hardening mainly by dissolved oxygen. The addition of Zr reduced the ion-irradiation hardening of steel to some extent mainly by suppressing the formation of dislocation loops. Meantime, as the FEM simulations revealed, the PDZ size underneath a conical indenter at the equivalent plastic strain εeq > 1.9% can reasonably estimate the density of GNDs. According to the strain gradient plasticity (SGP) theory, the densities of GNDs at an indentation depth of 200 – 300 nm were slightly higher in the case of the 50 dpa/nominal than the unirradiated, in spite of similarity for the two FeCrAl(Zr)-ODS steels.

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

Post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ∼10–13at. % Cr, ∼10–12 at. % Al, ∼1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a⟨100⟩ type loops is somewhat higher than the a/2⟨111⟩ type loops. The α′ precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α′ precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α′ precipitates on the dislocation loop density is discussed, and the presence of α′ appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α′ precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α′. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

Experiments and/or crystal plasticity finite element modeling of the mechanical properties of pristine and irradiated tungsten single crystal

Tungsten is used for vital applications in nuclear industry, particularly in experimental fusion reactors like ITER. Experimental evaluation of the irradiation-induced hardening is usually carried out by indentation tests. However, accurate prediction of irradiation hardening by indentation simulations faces two challenges: (1) compression data of single-crystal tungsten for indentation model calibration are lacking, since most available data are from tension tests, but tungsten exhibits a tension-compression asymmetry; (2) the irradiation hardening model for tungsten needs to be improved by incorporating more accurate irradiation defect strengthening mechanisms and defect density evolution laws that are quantitatively extracted from the lower scale modeling such as dislocation dynamics (DD) and molecular dynamics (MD). In this work, we present a comprehensive solution for the neutron irradiation hardening prediction of tungsten based on the experiments on pristine single-crystal tungsten and lower scale simulations. First, we conduct compression and Vickers indentation experiments to calibrate and validate the model for indentation simulations, and extend the model with the irradiation defect hardening law and the defect density evolution law informed by DD and MD simulations. Then, we use the irradiation hardening results from the literature to benchmark the model for its effectiveness, and find that the predicted hardness increases match well with the experimental results for different doses. Moreover, we discuss the drawback to the widely-used conventional hardening models, and find that these models tend to overestimate the defect strengthening coefficients.