Candidate Structural Materials Research Articles

Effect of cold work deformationon irradiation hardening of vanadium alloys

Vanadium alloys are regarded as promising candidate structural materials for the advanced blanket concept in fusion reactors due to their low activation, good high-temperature strength and, in particular, their compatibility with liquid lithium. In the present work, six kinds of V–5Cr–5Ti alloys under heavy cold work with deformation amounts of 40%, 60% and 80%, and/or subsequent annealing were investigated. Irradiation damage of 0.1, 0.3 and 0.5 dpa was introduced in both specimens using 352.8 MeV Fe ions at 100 °C. Electron backscattered diffraction and transmission electron microscopy (TEM) were used to investigate pre-irradiation microstructures such as grains, dislocations, precipitates and bubbles. X-ray diffraction was used to evaluate the pre-existing dislocation density and TEM was used to image the irradiation defects. The change in hardness was evaluated using micro-hardness tests. Before irradiation, the hardness increased with the increasing deformation amount but decreased after subsequent annealing. Dislocation cells turning into sub-grains with low-angle boundaries were observed, while the deformation amount reached 80% in cold-worked specimens. After irradiation, hardening was observed in all specimens and at all irradiation doses, and a power-law relation was observed in dose-dependent hardening. The effect of the initial microstructure on irradiation hardening was discussed in terms of the sink strength while ignoring grains and precipitates due to their large size. Pre-existing bubbles could effectively reduce irradiation hardening compared with previous results. Meanwhile, with the increasing sink strength of dislocations, hardening decreased in a different manner in cold-worked and annealed specimens. The irradiation defects in some specimens were investigated to clarify the inherent mechanism in the relationship between the initial microstructures and irradiation hardening.

Microstructure and mechanical property of novel nanoparticles strengthened AlCrCuFeNi dual-phase high entropy alloy

High entropy alloys are considered to be the best candidates for structural materials because of outstanding comprehensive properties resulted from their unique microstructure. Clarifying the microstructural evolution and controlling the content and morphology of the second phase are of great significance in the design of high-performance HEAs. Along this line of thinking, we report a research of adding little Mo(0–0.5 at%) to a novel Co-free AlCrCuFeNi alloy to tailor microstructure by introducing large lattice distortion and precipitation strengthening to achieve excellent comprehensive properties. The AlCrCuFeNi alloy exhibits BCC+FCC dual-phases, typical solid solution structures (B2 matrix, A2 flap-like and Cu-rich structure) and moderate mechanical properties. An ultrahigh hardness of 643 HV was achieved at the composition of Mo reaches 0.5 at%, which is approximately 1.5 times higher than AlCrCuFeNi alloy. A large amount of nano-sized σ precipitates with various sizes and morphologies were formed when the content of Mo was more than 0.3 at%. Most importantly, the Mo0.3 alloy exhibits an increase of ~18% in yield strength and an increase of ~30% in fracture strength accompanied by a reasonable plasticity which possesses potential application prospects in industrial areas. The solid solution strengthening and the formation of σ phase are the main factors contributing to the strengthening.

Mechanical characterisation of V-4Cr-4Ti alloy: Tensile tests under high energy synchrotron diffraction

Vanadium base alloys represent potentially promising candidate structural materials for use in nuclear fusion reactors due to vanadium's low activity, high thermal strength, and good swelling resistance. In this work, the mechanical properties of the current frontrunner vanadium base alloy, V-4Cr-4Ti, have been interrogated using in-situ high energy X-ray diffraction (XRD) tensile testing at varying temperatures. The single crystal elastic constants of the samples were determined from the in-situ XRD data and used to evaluate results from density functional theory calculations. Polycrystalline elastic properties and Zener anisotropy were calculated from the single crystal elastic constants produced, revealing the effect of elevated temperature on the alloy's elastic properties. These results characterise important thermomechanical properties, valuable in mechanical modelling, that will allow further and improved analysis of the structural suitability of V-4Cr-4Ti ahead of alloy adoption in the mainstream.

Ultrastrong and ductile additively manufactured precipitation-hardening medium-entropy alloy at ambient and cryogenic temperatures

Strong and ductile precipitation-hardening face-centered cubic medium-entropy alloys (MEAs) are potential structural material candidates for cryogenic applications, which, however, are rarely reported in the field of additive manufacturing. In this work, we develop a high-performance (CoCrNi)94Al3Ti3 MEA via additive manufacturing and age hardening. Superior tensile strength-ductility combinations, in comparison to other reported additively manufactured face-centered cubic metals, are achieved at both ambient (293 K) and cryogenic (103 K) temperatures. This is attributed to the hierarchical microstructure with a high degree of heterogeneity in terms of grain size, cellular substructure and characteristics of nanoprecipitates (L12 phase). Such microstructure feature leads to a higher ambient-temperature yield strength (921.1 MPa) and ultimate tensile strength (1346.4 MPa) than the as-printed sample that possesses a homogeneous microstructure. Moreover, the precipitate shearing mechanism, hetero-deformation induced hardening effect and deformation-induced stacking faults-based substructure evolution jointly result in a high and persistent strain hardening ability, which ensures a high ductility (27.2% at ambient temperature). The testing at the cryogenic temperature promotes the efficiency of hetero-deformation induced hardening and the formation of stacking faults, leading to an excellent strength-ductility combination (tensile strength 1702.9 MPa and ductility 25.4%). However, unlike the simultaneous increment of strength and ductility for the as-printed sample when the testing temperature decreases from 293 K to 103 K, the ductility of the age-hardened sample at different temperatures only changes slightly. This feature is related to the severe strain/stress concentrations developed within the heterogeneous microstructure of the age-hardened sample at 103 K. Our approach of introducing coherent nanoprecipitates in the additively manufactured microstructure provides a new insight into the development of high-performance MEAs for cryogenic applications.

Development of reduced activation ferritic/martensitic steels in China

Reduced-activation ferritic-martensitic (RAFM) steel is considered as the primary candidate structural material for blankets of ITER, CFETR and DEMO. Several RAFMs have been developed in China (CN-RAFMs) such as CLAM, CLF-1, modified RAFMs and oxide dispersion strengthened RAFMs (ODS-RAFMs). The mechanical properties, irradiation behaviors, additive manufacturing and joining technologies of structural materials have been comprehensively studied. Qualifications of CN-RAFMs are on-going for ITER-TBM fabrication, with emphasis on material property databases and standardization. A new neutron irradiation program is being carried out for a series of fusion materials e.g. RAFMs and ODS-RAFMs with doses up to 40 dpa. Joining technologies (e.g. electron beam welding, tungsten inert gas welding, laser beam welding and diffusion welding, etc.) and advanced additive manufacturing are performed to overcome the big challenges of the blanket fabrication before ITER, CFETR and DEMO. The databases and standards for CLAM and CLF-1 were established respectively, and the specification and qualification of them are underway for ITER-TBM.In this paper, the latest development and strategy on fusion energy in China are briefly introduced, and the progresses of R&D activities on CN-RAFMs aiming for the engineering applications are reviewed.

Effects of local chemical ordering on defect evolution in NiFe concentrated solid solution alloy

Single-phase concentrated solid solution alloys (SP-CSAs), including high-entropy alloys (HEAs), are promising candidates for structural materials in nuclear reactors. Recent studies have demonstrated that local chemical ordering (LCO) is an important structural feature in these alloys, which can significantly affect defect evolution. The presence of LCO leads to elemental rearrangement in CSAs and results in distinct interfaces separating the ordered and disordered phases. In this work, we study the role of LCO and order/disorder interface on defect evolution in NiFe CSAs. Our results indicate that the interface can act as defect sinks in the defect recombination stage. At the diffusion stage, defect evolution is strongly affected by LCO due to chemically biased diffusion in CSAs and preferential diffusion pathways of defect clusters. We further show that the degree of LCO and the dimensions of the LCO domains play significant roles in defect clustering and annihilation behavior. These results provide fundamental insight into the influence of LCO on defect evolution in CSAs and pave ways for designing highly irradiation-tolerant structural materials for nuclear applications.

Compatibility of Different Commercial Alloys in High-Temperature, Supercritical Carbon Dioxide.

In this work, the compatibility and long-term integrity of candidate structural materials, including the austenitic stainless steel 316NG, the Fe-Ni-based alloy 800H, and the Ni-based alloy 625, were tested in high-temperature and high-pressure SCO2. The exposure time was up to 3000 h. The results showed that the corrosion kinetics approximately followed a near-cubic law for 316NG and 800H. After 3000 h exposure, all oxide layers, mainly composed of Cr2O3, were continuous, compact, and protective, and their thicknesses were about 21~45 nm, 64~88 nm, and 34~43 nm, respectively. In the case of carburization, dark spots corresponding carbon deposition were observed on the surface and a little enriched in the underside of the oxide for 800H. Moreover, the enrichment of trace elements was found at the oxide/substrate interface through GDOES and TEM analyses, i.e., the enrichment of Mn and Si for 316NG, the enrichment of Mn, Si, Al, and Ti for 800H, and the enrichment of Ti and Al for alloy 625.

A potential candidate structural material for molten salt reactor: ODS nickel-based alloy

The commercialisation of molten salts reactors (MSRs) is hindered by the lack of structural materials capable of withstanding the corrosive environment therein. To address this problem, we herein prepared 1 wt% Y2O3 dispersion–strengthened NiMo-based alloys using powder metallurgy and evaluated their potential as structural materials for MSRs based on their mechanical properties, He swelling behaviour, and molten salt corrosion resistance. In view of the strengthening provided by homogenously dispersed Y2O3 particles, all NiMo-Y2O3 samples exhibited ultimate tensile strengths and yield strengths exceeding those of the Hastelloy N alloy, a state-of-the-art structural material for MSRs. Moreover, the volume fraction of He bubbles in the NiMo-Y2O3 samples (∼0.3%) was lower than that in the Hastelloy N alloy (0.58%), which showed that the introduction of Y2O3 nanoparticles effectively inhibited He swelling. All NiMo-Y2O3 samples showed excellent resistance to molten salt corrosion (as reflected by the absence of obvious holes therein), thus holding great promise for the development of irradiation- and molten salt corrosion–resistant structural materials for high-temperature MSRs.

Effect of Silicon on Dynamic/Static Corrosion Resistance of T91 in Lead-Bismuth Eutectic at 550 °C.

The 9–12% Cr ferritic–martensitic heat-resistant steel is the main candidate structural material for the Lead-cooled Faster Reactor. The lower Gibbs free energy change of Si oxide can promote the formation of a stable oxide layer, which can improve the corrosion resistance of the material. Therefore, it is of great significance to study the effect of silicon (Si) on the corrosion resistance of T91 steel in lead–bismuth eutectic (LBE). The corrosion resistance of T91 steel with Si contents of 0.5 wt.%, 1.3 wt.%, and 2.0 wt.%, both in dynamic and static LBE at 550 °C, was investigated. The microstructure was analyzed by X-ray diffraction (XRD), scanning electron microscope (SEM) and transmission electron microscope (TEM), while the oxide films were characterized by electron probe microanalysis (EPMA). Results show that the addition of Si is conducive to improving the corrosion resistance of T91 steel in LBE. T91 steel with high Si content has a thinner and more stable oxide film. The change of Si content can change the contact angle between the steel and LBE, and the contact angle is the largest when the Si content is 1.3 wt.%. The Si-rich oxide layer is usually located in the inner oxide layer, which promotes the formation of a Cr oxide layer located in the internal oxidation zone (IOZ). Si will not enter the precipitated phase, but only change the ferrite content. The oxidation model of T91 steel containing Si in LBE was also proposed.

Enhanced fatigue property by fabricating a gradient nanostructured surface layer in a reduced-activation steel

Reduced activation ferritic/martensitic (RAFM) steels have been selected as candidate structural materials for future advanced nuclear power systems. In the present work, the influence of a gradient nanograined surface layer on the fatigue properties of RAFM steels was studied. A gradient nanostructured (GNS) surface layer with a thickness of ∼85 ​μm was prepared on RAFM steel utilizing surface mechanical rolling treatment (SMRT). The mean grain size was approximate 43 ​nm at the topmost surface and increased gradually with depth. The results of the stress-controlled tension-compression fatigue experiments showed that the fatigue life enhanced approximately 6 times in the SMRT samples compared to the corresponding base metal counterparts. The relationship between the applied stress amplitude and the fatigue lifetime, and the fracture morphology showed that the surface strengthening and strain delocalization were caused by GNS, which suppressed surface crack initiation process, and hence the fatigue properties of RAFM steels improved. In addition, the deformation compatibility in GNS and coarse-grained boundaries leading to more dislocation interactions and accumulation during the cyclic process, also plays a crucial role in enhancing the fatigue properties of RAFM steel.

Predicting the Irradiation Swelling of Austenitic and Ferritic/Martensitic Steels, Based on the Coupled Model of Machine Learning and Rate Theory

As nuclear structural materials, austenitic and ferritic/martensitic (F/M) steels will face inevitable irradiation swelling when fulfilling a role in nuclear reactors, especially under high-dose irradiation. For this work, a coupled machine learning rate theory (ML-RT) model for the swelling of austenitic and F/M steels was developed. In this model, ML was introduced to predict the steady-state irradiation swelling onset dose (Donset), while the improved RT was developed to simulate the swelling behavior after the incubation period. More than 200 series of data on the Donset of different structures of steel were collected for the ML prediction. The coefficient of determination (R) of the results in ML is more than 0.9. In the RT, the evolutions of the dislocation loop and void were described and calculated rather than using the fitting parameters. Cascade efficiency was employed to describe the cascade process. The coupled ML-RT model was verified with the swelling data from neutron irradiation experiments for various steels. The theoretical results of the swelling peak temperatures and swelling behavior are more accurate and reasonable, compared with those from the previous RT model. Using the ML-RT model, the swelling performance of CLAM steel under neutron irradiation of up to 180 dpa was predicted. The differences between the swelling performance of austenitic steels and F/M steels were analyzed and the differences were mainly associated with the bias. These results will be helpful for evaluating the neutron irradiation swelling behavior of candidate structural materials.

Fabrication of Oxide‐Dispersion‐Strengthened Ferritic Alloys by Mechanical Alloying Using Pre‐Alloyed Powder

Mechanical alloying (MA) is widely used to produce oxide‐dispersion‐strengthened (ODS) steels, which are potential candidate structural materials for the next‐generation fusion reactors. Here in this study, two kinds of ODS ferritic alloys, labeled A and B, are prepared through MA, followed by spark plasma sintering (SPS) and annealing. To prepare Alloy A, Fe powder, YH2 powder, and TiO2 powder are mixed for MA. To prepare Alloy B, a pre‐alloyed Fe–Cr–W–Y powder is first obtained by gas‐atomization, after which TiO2 powder is added for MA. Compared with the traditional MA method, where the highly stabilized Y2O3 powder is usually selected, the use of TiO2 powder successfully reduces the milling time since it should be more easily decomposed and dissolved into the ferrite matrix. Some TiO2–Y2TiO5 duplex oxide particles of about 50 nm are found to generate through diffusion‐induced in situ reactions in Alloy A. In Alloy B, in contrast, a high number density of well‐distributed Y2O3 particles of about 20 nm and ultrafine Y2Ti2O7 particles within 10 nm are observed. These nanoparticles maintain good coherent interfaces with the matrix, suggesting the precipitation mechanism. Herein, a new route for the efficient fabrication of ODS alloys is provided.

Improving the exploration of vacancy evolution in P92 alloy under Fe ion irradiation using positron annihilation

P92 alloy, a candidate structural material for supercritical water reactors, was irradiated with 1 MeV Fe ions at 1, 3, and 7 dpa at room temperature, 300, 450, and 550°C. Slow positron beam Doppler broadening spectroscopy was used to characterize the evolution of the open-volume defects. Irradiation of the Fe9Cr alloy at 300°C induced an unexcepted phenomenon in which the S parameter decreased with an increasing irradiation dose; the phenomenon was ascribed to the pinning effect of Cr on dislocation loops, which promoted the recovery of vacancies. Whereas the reduction in the S parameter was generally inhibited in the P92 alloy. These distinctions are related to the evolution of vacancies. We show that the migration of vacancies hindered by C atoms is mainly related to carbon–vacancy complexes in the P92 alloy at approximately 300°C. As the irradiation temperature increased, the complexes gradually dissociated, and the vacancies were captured and annihilated with interstitial atoms. A better understanding of the evolution of vacancies is helpful for investigating irradiation swelling.

Hierarchical structure and deformation behavior of a novel multicomponent β titanium alloy with ultrahigh strength

Based on the general [Mo] equivalent criterion and d-electron orbital theory, a new ultrahigh-strength β titanium alloy with eight major elements (Ti-4.5Al-6.5Mo-2Cr-2.6Nb-2Zr-2Sn-1V, TB17) for industrial applications was developed. An ingot of five tons was successfully melted by thrice vacuum consumable arc melting. The microstructure and elements partitioning of different conditions were investigated systematically. The results suggest that the hierarchical structures of micro-scale first α phase (α f ), nano-scale secondary α phase (α s ), and ultrafine FCC substructures can be tailored by solution plus aging (STA) heat treatment. The lateral and epitaxial growth of α f phase promotes the HCP-α to FCC substructure transformation with the help of elements partitioning during the aging process. Moreover, the element V, generally regarded as β stabilizer, is found to mainly concentrate in the Al-rich α f phase in this study probably due to its relatively lower content and the strong bonding energy of Al-V. The hierarchical structure has a strong interaction with dislocations, which contributes to achieve a superhigh strength of 1376 MPa. In addition, the plastic strain is partitioned in the multi-scale precipitates (such as the α and FCC substructures) and β matrix, resulting in a considerable plasticity. TEM observation demonstrates that high density entangled dislocations at interfaces and mechanical twins exist in the STA sample after tensile test. It can be deduced that both dislocation slipping and twinning mechanisms are present in this alloy. Therefore, TB17 alloy can serve as an excellent candidate for structural materials on aircrafts that require high strength and lightweight. • A novel β titanium alloy with ultrahigh strength of 1376 MPa was developed. • The multi-scale hierarchical structures were tailored by heat treatment. • The transformation from HCP-α to FCC-substructure was observed. • The V element mainly concentrated in the micro-scale α lath.

Electrochemical Noise Response of Cr2Nb Powders Applying Mechanical Alloying

Cr2Nb alloys are potential candidates for high-temperature structural materials. The influence of different mechanical alloying parameters (milling time) and sintering processes were studied. After mechanical alloying and observation by scanning electron microscope (SEM), nano powders were characterized and then sintered by spark plasma sintering (SPS). Electrochemical noise (EN) tests were also conducted in order to study the electrochemical behavior. From the current experimental results, it was revealed that ball milling times up to 20 h may explain the influence of Nb–Cr alloys and its association to the Laves phase and corrosion behavior. These insights aimed at improving the samples’ predicted behavior before spending time and resources at high-temperature industrial processes.

Microstructure Characterization and Small Punch Test Analysis in Nickel-Based Alloy 617 by High Energy Neon Implantation

Nickel-based alloys are good candidate structural materials for ultra-high temperature gas-cooled reactors due to their excellent mechanical properties under high-temperature conditions. The operating environment in the compact high-temperature gas-cooled reactor proposes more stringent requirements with higher displacement and temperatures higher than 700 °C. The irradiation resistance of nickel-based alloy 617 is evaluated by 120 MeV neon ion irradiation. The neon ion fluence is set with reference to the number of helium atoms generated by the reaction of (n, α) in nickel-based alloy under actual service conditions. With the application of an energy gradient degrader, the irradiation damage area is more uniform with increasing depth. The neon ion irradiation causes no significant surface damage to alloy 617 while the evolution of defects such as inherent dislocation lines is affected, thus leading to the deterioration of mechanical properties. By using the small punch test to analyze the irradiation effects, it is found that the neon ion irradiation results in the tendency of irradiation softening and does not lead to changes in the basic fracture characteristics under different annealing conditions. After annealing at 700 and 800 °C for 2 h, the irradiation embrittlement trend is intensified, with the fracture characteristic values of irradiated samples showing a more obvious decreasing trend.

Fe+ ion irradiation effects in Fe-10at%Cr films irradiated at 300 °C

Fe-Cr alloys constitute the model systems for the investigation of radiation damage effects in ferritic-martensitic steels which are candidate structural materials for fusion reactors. In the current study Fe-10at%Cr alloy films of 70 nm thickness were irradiated by 490 keV Fe + ions at 300 °C at doses ranging from 0.5 up to 20 displacements per atom (dpa). The Fe + ion energy chosen corresponds to the energy of primary Fe(Cr) knock-on atoms from 14 MeV neutrons. The irradiation effects were investigated employing X-ray diffraction and X-ray and polarized neutron reflectivity. The irradiation produced dose dependent: a) lattice constant increase, b) grain size growth and c) Cr depletion in the matrix. These changes occur largely up to 4 dpa and afterwards the system attains a dynamic equilibrium.

Revealing the role of W in the microstructure, hardness and weldabilites of new developed Ni-xW-6Cr alloys

Ni-xW-6Cr alloy is candidate structural material for advanced high temperature reactors due to the excellent high temperature mechanical properties and corrosion resistance to molten salts. However, high tungsten content adversely affects the weldability. In this work, Variable restraint test was carried out on Ni–W (5%∼35%)-6Cr alloys, microstructure, mechanical properties and weldability were studied. Results show that the increase of W content significantly improved the lattice size, grain size and hardness of the alloy. The hardness of weld metal is obviously weaker than that of base metal after welding. Moreover, higher W content decreased the weldabilities of Ni-xW-6Cr alloy. The solidification cracks occurred in the weld metal when the tungsten content exceeds 20 wt%. The stress concentration tended to occur at the solidification grain boundaries and promoted cracks to initiate and propagate along the solidification grain boundaries.

Dense additive-free bulk boron nitride ceramics developed by self-densification of borazine

Boron nitride (BN) ceramics are promising candidates for high-temperature structural and functional materials. However, it is difficult to sinter dense additive-free bulk BN monoliths because of the covalent bond and flake structure. Here, we report dense bulk BN with relative density higher than 95% without sintering additives achieved via a self-densifying approach of borazine within a wide temperature range from 800 °C to 1800 °C. A polyborazylene with a controlled degree of cross-linking was synthesized and press-molded into shaped green bodies, which were pressureless pyrolyzed into porous frameworks and then densified through repeated borazine infiltration and pyrolysis method. The microstructural and crystalline evolution of the bulk BN ceramics, as well as the corresponded mechanical, thermal and dielectric properties were investigated.

Effects of sintering temperature and Y2O3/Ti addition on microstructure and hardness of ODS-AlCrFeNi HEAs

Oxide dispersion strengthened (ODS) alloys are leading candidate structural materials for advanced nuclear reactor application due to their excellent high strength and radiation tolerance. In this paper, ODS AlCrFeNi high entropy alloys (HEAs) were synthesized via mechanical alloying (MA) and spark plasma sintering (SPS) to study the effects of sintering temperature (from 1050 °C to 1250 °C) and different Y2O3/Ti addition amounts on the microstructure and hardness. The microstructure was characterized by XRD, SEM, EBSD and HRTEM, and the micro-hardness was tested. The results demonstrated that after MA and SPS, spinodal decomposition occurs in the matrix, in which Fe/Cr enriched disordered BCC phase and Ni/Al enriched ordered B2 phase are formed. At the higher sintering temperature (1250 °C), additional FCC phase appears in 0.3 YT, which is contributed to the formation of higher volume fraction of Al2O3 at powder/grain boundaries so that spinodal decomposition is partly suppressed. At the lower sintering temperature (1100 °C), FCC phase could be formed when Y2O3+Ti amount is lower (0.6 wt%, 1.0 wt%). The hardness of ODS-AlCrFeNi is notably improved when Y2O3 and Ti are added in the same weight percent which is up to 1 wt% or above. The average grain size of 3.0 YT alloy (3 wt% Y2O3+3 wt% Ti) is the smallest, which is 240 nm. The average size of precipitates in 3.0 YT alloy is 8.7 nm, while the number density is the highest (3.3 × 1022/m3). The addition of Y2O3/Ti can modify the types of nanoscale oxides in the ODS-AlCrFeNi. The precipitates in 0Y alloy (without Y2O3 and Ti addition) are Al-rich oxides, while the precipitates observed in 3.0 YT alloy include AlYO3, Al2Y4O9, Y2TiO5, YTiO3, Al2TiO5 and Al2O3. The grain boundary strengthening and precipitation strengthening are dominated in 3.0 YT alloy leading to the highest hardness increment compared with 0Y alloy and the hardness of 3.0 YT alloy reaches 751 HV.