Zirconium-based Alloys Research Articles

Optimizing Manufacturing of Zr–Cu–AI–NI Metallic Glasses via Laser Metal Deposition

Recently, it was discovered that the cutting-edge technique known as laser powder bed fusion (LPBF) is the best way to produce Zr-based bulk glasses made of metal (BMGs). While LPBF gives greater versatility, current state-of-the-art production techniques like copper mold casting and arc-melting have limits when it comes to implementing complicated designs. Furthermore, LPBF enables a delicate balance to be struck between producing intricate characteristics and sustaining suitable temperatures all through the whole operation. Because of its exceptional features and practical availability, this research focuses on optimizing the process variables for a specific Zr-based alloy, AMZ4, which is produced by additive manufacturing in order to optimize both its mechanical and thermal characteristics. Belonging to the class of zirconium-based alloys known as bulk metallic glasses (BMG), Zr57Cu15Ni10AI5 (or Vit-106) has an excellent glass-forming ability and shows great promise. By casting, a BMG alloy may be transformed into workpieces that are about one centimeter in size in all three dimensions. Nevertheless, crystallization is induced when the cast size is further increased since it reduces the cooling rate. By building a workpiece from many melt sections with the cooling rate maintained above the critical one, selective laser melt (SLM) is an established technique for overcoming size restrictions for BMGs. Partially crystallized BMG is now an issue with SLM-obtained components. The effect of SLM process variables on partial crystallization is investigated in this paper. You may regulate the size and intensity of the inclusion by altering the speed of the laser scanning. Microhardness and wear resistance may be improved by incorporating submicron crystalline inclusions into the amorphous matrix by SLM.

Structure and Corrosion Behavior of Multiphase Intermetallic ZrCu-Based Alloys.

Zirconium-based alloys are highly regarded by the research community for their exceptional corrosion resistance, thermal stability, and mechanical properties. In our work, we investigated two newly developed alloys, Zr42.42Cu41.18Al9.35Ag7.05 and Zr46.81Cu35.44Al10.09Ag7.66, in the form of ingots and ribbons. In the course of our investigation, we conducted a comprehensive structural and thermal analysis. In addition, an examination of the corrosion activity encompassing electrochemical studies and an analysis of the corrosion mechanisms was carried out. To further evaluate the performance of the materials, tests of their mechanical properties were performed, including microhardness and resistance to abrasive wear. Structural analysis showed that both alloys studied had a multiphase, crystalline structure with intermetallic phases. The samples in the form of ribbons showed improved corrosion resistance compared to that of the ingots. The ingot containing a higher content of copper Zr42.42Cu41.18Al9.35Ag7.05 was characterized by better corrosion resistance, while showing lower average hardness and a higher degree of abrasive wear based on SEM observations after pin-on-disc tests.

Microstructure and mechanical properties evolution of the Cr coatings on SiC substrates under He irradiation

Environment coatings are frequently considered, to improve the corrosion resistance of silicon carbide (SiC) in pressurized water reactor (PWR) and boiling water reactor (BWR) surroundings. Chromium (Cr) is one of the main candidate materials for coatings on zirconium-based alloys for accident-tolerant fuel (AFT) cladding. In this work, the magnetron sputtered Cr coatings on SiC substrates were irradiated by the 500 keV He+ ions at room temperature. Moreover, the microstructural characterization was conducted on the irradiated specimens by using scanning electron microscope (SEM), and transmission electron microscope (TEM). The scratch test showed that the coating-substrate adhesion decreased from 41 N to 28 N (∼31.7%) before and after the irradiation, maybe due to that the irradiation produced tensile strain in the coating and mesophase along the coating-substrate interface. It revealed that irradiation accelerated the migration and aggregation of C elements on the surface of the coating. After the 800 °C annealing for 20 min, massive Cr atoms diffused into the SiC matrix and reacted with Cr to form a chromium silicide interlayer. Some cracks were observed at the interface, being attributed to the Kirkendall effect. Therefore, it is very necessary to avoid the formation of chromium silicide interlayer and decrease swelling mismatch between the metal coatings and the SiC substrate for the actual application in nuclear systems.

Lithium stabilization of amorphous ZrO2

Small modular reactors (SMRs) are a key option to aid the worldwide net zero targets for carbon emissions. Some pressurised water reactors aim to operate with a boron-free coolant chemistry for simplification in plant design. In the absence of boron, Li has been found to accelerate the corrosion of the zirconium-based alloy fuel cladding under certain conditions and concentrations within pressurised water reactors (PWRs). The cause of the accelerated corrosion has yet to be identified. This work identifies the potential for amorphous ZrO2 phase stabilization by lithium that will potentially have an impact on the passive oxide's grain boundary structure. Higher lithium concentrations are found to stabilise the amorphous phase's stability before re-crystallisation to higher temperatures. Chemical assessment has also shown the lithium phase to be soluble in water, indicating a potential mechanism for lithium to alter grain boundaries and increase the pathways for further oxidation of the underlying zirconium material.

Post-quench ductility limits of coated ATF with various zirconium-based alloys and coating designs

Post Quench Ductility (PQD) of various Cr-coated cladding designs with respect to base cladding materials, Cr coating methods, and its thickness were investigated to explore Emergency Core Cooling System (ECCS) limits. Both inner wall and outer wall of coated (Cr and Cr/CrN) claddings were steam oxidized at ∼1204 °C and then water quenched. The post-oxidized specimens were ring compressed to attain offset strains based on which ductility was assessed in compliance with the U.S. Nuclear Regulatory Commission (U.S. NRC)’s protocols. Weight gain-based Equivalent Cladding Reacted (ECR) limits of coated specimens (Cr and Cr/CrN) are consistently lower than those of the base cladding materials due to the early load drop under Ring Compression Test (RCT). Yet, time needed to reach the ECR limit is still increased for coated specimens because it effectively undergoes single side oxidation with protective coating. Cracks imitated from ZrCr2 are primarily responsible for the early major load drop observed for tested coating thicknesses (8.9 µm and 18.9 µm), and hence reduced ECR limits compared to bare Zircaloy. The cracks from ZrCr2 was promoted by an increased oxygen concentration of the interfacing Zr matrix owing to the diffusion of oxygen through the coating. The thicker coating reduces this oxygen level, thereby delaying the load drop from ZrCr2 during RCT and increases the ECR limit. The maximum attainable ECR limit of coated ECR limits is affected by the base Zircaloy as it gives as the ECR ceiling from which ECR reduction for coated cladding occurs. For the tested base cladding materials (Opt. ZIRLO™ and HANA-6), ECR 19 %, which is the limit for Cr-coated (8.9 µm) HANA-6, may serve as the lower envelope limit. Hence, ECR 19 % (single-side ECR) can conservatively serve as the conservative, yet non-design specific, Cr-coated ECR limit for the most of the modern base Zircaloy materials (Opt. ZIRLO™ and HANA-6).

Phase Composition, Microstructure and Mechanical Properties of Zr57Cu15Ni10Nb5 Alloy Obtained by Selective Laser Melting

Zr57Cu15Ni10Nb5 (more known as Vit-106) is a promising zirconium-based alloy with a high glass-forming ability, and belongs to the so-called bulk metallic glasses (BMG). Workpieces with a size of around one centimeter in all three dimensions can be obtained from a BMG alloy by casting. However, further increasing the cast size decreases the cooling rate and thus induces crystallization. Selective laser melting (SLM) is a well-known technique to overcome size limitations for BMGs because a workpiece is built by the addition of multiple melt portions in which the cooling rate is kept above the critical one. Currently, BMG parts obtained by SLM suffer from partial crystallization. The present work studies the influence of SLM process parameters on the partial crystallization of Vit-106 by metallography and the influence of the microstructure on mechanical properties by microhardness and wear resistance testing. Submicron crystalline inclusions are observed in an amorphous matrix of a Vit-106 alloy obtained by SLM. The size and the concentration of the inclusions can be controlled by varying the laser scanning speed. It is shown that submicron crystalline inclusions formed in the amorphous matrix during SLM can favorably affect microhardness and wear resistance.

Microstructure evolution of the Zr-4 alloy joints diffusion bonded with pure titanium interlayer and its influence on joint properties

Zirconium-based alloy is widely applied as the fuel cladding material. Its weld needs high precision and small deformation, so the low-temperature bonding is a crucial issue to be addressed. In this study, we used pure titanium interlayer to join Zr-4 alloys by diffusion bonding at low temperature. A sound joint with pure titanium interlayer was bonded at 680 °C, which was 70 °C lower than that of the direct bonding. The effect of welding parameters on the microstructure and phase composition was investigated by scanning electron microscope (SEM) with energy dispersive spectroscopy (EDS), and transmission electron microscope (TEM). Four solid solutions, (hexagonal close-packed) hcp-Ti0.92Zr0.08, hcp-Ti0.7Zr0.3, hcp-Ti0.5Zr0.5, and (body-centered cubic) bcc-Ti0.5Zr0.5, were identified forming at the Ti/Zr interface by EDS and TEM analysis. The nanohardness and Young’s modulus were measured by nanoindentation. The highest and lowest nanohardness values across the joint were 5.65 GPa for the hcp-Ti0.7Zr0.3 phase, and 4.09 GPa for the bcc-Ti0.5Zr0.5 phase, respectively. The fracture initiated in the bcc-Ti0.5Zr0.5 phase, prolonged at the Ti/Zr interface because of incoordination in the deformation process of two phases, and then the transverse tearing occurred in the hcp layer. The shear strength initially increased and then decreased as the increasing parameters. The maximum shear strength reached 255 MPa of the joint bonded at 750 °C for 60 min under 15 MPa. Besides, the formation mechanism of the Ti/Zr interface was revealed for the Zr-4 alloy joint with pure titanium interlayer.

Peculiarities of interaction between oxygen and solute/vacancy in commercial-type zirconium-based alloys: A first-principles study

Unveiling the interaction between dissolved oxygen and solutes/vacancies during the initial stages of oxide layer formation at microscale is significant for the peculiarities of Zr-based alloys oxidation exposed in PWRs or LWRs. We perform ab-initio calculations of the diluted zirconium-based alloys with dissolved Nb and Sn solutes, vacancy and oxygen atom and discuss the solute-oxygen and vacancy-oxygen binding. Our analysis shows that with increasing content of Nb solutes the attractive interaction between niobium and oxygen becomes much stronger, which results in a decrease in corrosion rate of these alloys at the bre-breakaway regime caused by diffusion of oxygen. An increase in content of Sn in an alloy results in enforce of the repulsive interaction between Sn atom and oxygen, which causing tin-enriched alloys are characterized by an increased corrosion rate. We conclude that the dissolved oxygen can be localized near Nb atom rather than near Sn atom if both impurities are second-nearest neighbors, the dissolved oxygen can capture single vacancy and small vacancy clusters in zirconium matrix. This study provides a deep insight into the details of oxygen segregation in zirconium-based alloys exploited in nuclear-power plants.

Simulation of the compression test of the Zr1Nb fuel cladding ring

Fuel cladding is a first protective barrier against the loss of fission products that must withstand extreme conditions, from normal operation to final and interim dry storage. This hostile environment results in mechanical and microstructural damage of cladding caused by different stress levels, temperature, corrosion, hydrogen pick up and other degradation processes further enhanced by radiation. For this reason, the integrity of the cladding is a critical issue. The aim of this work is to simulate a ring compression test to evaluate the stress-strain behavior and hoop fracture properties of a zirconium-based alloy with niobium, which was chosen because it is widely used as fuel cladding in light water nuclear reactors.

Effect of solute Nb and Sn on self-interstitial atom defect in zirconium-based alloys by first-principles calculations

Zirconium based alloys are viewed as one of the most important fuel cladding materials, and alloy elements play a crucial role in improving its irradiation resistance, especially Nb and Sn. In present work, self-interstitial atom formation energy, binding energy between dopant Nb/Sn and the self-interstitial atom, and electronic structure are systematically investigated with the density functional theory. The diffusion barriers of the self-interstitial atom under the effect of Nb/Sn are calculated with the climbing image nudged elastic band (CI-NEB) method. Depending on the calculation results, the most stable configuration is BT and BS after doping Nb and Sn atom, respectively, different from BO for the Zr crystal without dopant. In the doped structure, the orbital hybridization of Nb atom and self-interstitial atom was obvious, while the orbital overlap of Sn atom with matrix atom and self-interstitial atom was less. Orbital hybridization of Nb atom and zirconium atoms mainly occurs between d orbits. After doping Nb, the diffusion barrier of the interstitial atoms increases, inhibits their diffusion or cluster growth. And the dopant Sn makes the formation of interstitial atom more difficult depending on the larger formation energy. The calculation results may explain the mechanism of improving its irradiation resistance by Nb and Sn doping.

Assessing Li accommodation at amorphous ZrO2 grain boundaries

Nuclear Pressurised Water Reactors (PWRs) use zirconium alloys as a fuel cladding, preventing the cooling water, at elevated pH using lithium hydroxide, from interacting with the fuel. Boron, as boric acid, is added to the coolant as a reactivity shim. Future reactor designs are considering removing soluble boron reactivity control to aid plant simplification. The presence of lithium in the absence of boron in the coolant has, however, been found to accelerate the corrosion of zirconium-based alloys under certain conditions and the mechanisms by which this occurs is under investigation. The ingress of lithium into the bulk oxide layer of zirconium alloy has been addressed in a previous study and was found to be unlikely. Here, atomistic simulations were used to produce Brouwer diagrams from which the solubility of lithium in amorphous structures representing complex grain boundaries have been predicted. The solubility of lithium in these amorphous structures is predicted to be high and will produce an elevated concentration of oxygen defects within the amorphous structure. This could offer a mode for transport of oxygen to the metal oxide interface and, potentially, offer a mechanism or part of a mechanism for observed lithium-accelerated corrosion of Zr-based alloys.

Cladding of nuclear fuel with iron based alloy: Penalty and solution

Replacing zirconium-based alloys fuel cladding by that of iron in nuclear fuel of Light Water Reactors (LWRs) needs increasing the fuel enrichment to achieve the same core cycle length. The penalty of using iron based alloys fuel cladding in LWRs can be overcome by direct recycling of nuclear spent fuel in CANDU reactors. The problems associated with using zirconium-based alloys fuel cladding in LWRs is not significant in the case of CANDU reactors. This is because CANDU reactors discharge the spent fuel with low burnup which is in the withstand tolerance of zirconium-based alloys and the moderator acts as an effective heat sink in case of severe accident.The case studied in this research is direct recycling of the spent fuel discharged from the European Advanced Pressurized Water Reactor (EU-APWR) in CANDU-6 reactor. The calculations using MCNPX computational code showed that replacing fuel clad from zircaloy-4 to stainless steel-348 requires increasing the fuel enrichment from 4.95 %w U-235 to 6.8 %w U-235 with an increase of 37%. After direct spent fuel recycling in CANDU-6, the total burnup of the fuels cladded with zircaloy-4 and with stainless steel-348 are about 70 and 96 GWd/t, respectively with an increase of about 37% in the case of the fuel cladded with stainless steel which compensates the increase of the enrichment.

Understanding the role of tetragonal phase in oxidation behaviour of different zirconium-based alloys in air and water vapour conditions – Phase fraction and stress field distribution

The paper investigates the influence of oxidation atmosphere on the corrosion behaviour after 7, 15, and 24 h of different zirconium-based materials, including pure metal, E110, and Zircaloy-2 alloys. The study utilises scanning electron microscopy to perform cross-sectional characterisation. Raman imaging was employed to determine the content and distribution of the tetragonal phase in the oxide scale depending on oxidation parameters. The obtained results were compared with previous thermogravimetric measurements and correlated with stress distribution calculated based on the shift of the selected tetragonal phase band. Additionally, the study reveals a relaxed tetragonal phase in a bulk monoclinic phase. A thorough analysis of Raman spectra allowed correlating spectral features with tetragonal phase stabilisation mechanisms for the tetragonal phase occurring at the metal/oxide interface and in the bulk monoclinic phase.

Experimental and modelling evidence for hydrogen trapping at a β-Nb second phase particle and Nb-rich nanoclusters in neutron-irradiated low Sn ZIRLO

Zirconium-based alloys used for fuel cladding in nuclear fission reactors are susceptible to hydrogen embrittlement during operation, but we currently lack the necessary mechanistic understanding of how hydrogen behaves in the materials during service to properly address this issue. Imaging the distribution of hydrogen within material microstructures is key to creating or validating models that predict the behaviour and influence of hydrogen on material properties, but is experimentally difficult. Studying hydrogen in zirconium-alloys is further complicated by the fact that the most common routes for preparing specimens for Transmission Electron Microscopy and Atom Probe Tomography (APT) analysis, electropolishing and focused ion beam (FIB) milling, are known to induce hydride formation. This introduces uncertainty as to whether the hydrogen distribution in the analysed specimen is actually representative of the entire sample a priori. Recent work has shown that this effect can be mitigated by performing the final specimen thinning stages at cryogenic temperatures. In this paper we use cryo-FIB to prepare APT specimens of neutron-irradiated low Sn ZIRLO, showing that hydrogen is trapped within a β-Nb SPP and at Nb-rich nanoclusters formed by exposure to neutron irradiation. We then use density functional theory calculations to explain these experimental observations. These results highlight the importance of including niobium-rich features in models used to predict hydrogen pick-up in zirconium alloys during service and delayed hydride cracking during storage.

Thermomechanical evaluation of ATF fuel rods regarding irradiation effects

Fuel clad is the main barrier against the release of radioactive materials. For decades zirconium-based alloys have been used as nuclear fuel clads. Fukushima accident proposed the idea of designing and developing accident tolerant fuels (ATF). This research tries to evaluate the effect of irradiation damage on ATF clads under operating condition of a pressurized water reactor. In this work, the SiC and FeCrAl clads are investigated as two candidates for replacement with conventional cladding, Zircaloy-4. The neutronics of a WWER1000 reactor is modeled based on the Monte Carlo method. The neutronic evaluation of the core with desired clads is performed over three years in different spatial points. Using calculated neutron spectrum, the SPECTER and SPECOMP codes are applied to evaluate the damage caused by the irradiation in the term of displacement per atom (DPA). The thermomechanical evaluation of the fuel rod considering cladding irradiation effects and burnup are performed by a developed thermomechanical code. Also, stress, strain, creep, fuel-clad contact, thickness of oxide layer and … are calculated and the results are compared considering irradiation effects.

Unveiling the characteristics of the residual point defects of collision cascade in Zr-xNb binary system: A molecular dynamics study

Zirconium-based alloys are commonly used as fuel cladding materials for light-water reactors. As an alloying element, niobium can further improve the performance of Zr alloys. However, the mechanism of the Nb effect on the irradiation damage (e.g., the irradiation-induced precipitation of Nb-rich clusters) of the alloy is still unclear and should resort to atomic-scale modeling. In this work, we unveiled the characteristic of the residual point defects produced at the end of the collision cascade caused by primary knock-on atoms (PKAs) in the Zr-xNb binary alloy through many molecular dynamics simulations. The effects of PKA directions (<0001> and <101¯0>), PKA energies (2 keV, 5 keV, 10 keV, 20 keV, 30 keV, and 50 keV), temperatures (300 K, 600 K, and 800 K), and the initial Nb contents in the alloy (cNb = 0 at.%, 0.5 at.%, 1 at.%, and 2 at.%) on the remaining Frenkel pairs, the displaced atoms, the composition of these point defects, the size and number of clusters, and the composition of the clusters with different-size have been studied. Several concepts were clarified from these simulation results. A special focus of this work is on the Nb proportion in the point defects and their clusters. It is worth emphasizing that the interstitial-Nb proportion in the interstitials and their clusters is far greater than the initial Nb content in the alloy, which can be the nucleate center of Nb-rich precipitation particles. This distinct feature may be solving the experimental puzzle of radiation-induced Nb precipitates/clusters of Nb-containing Zr alloys.

B2 to ordered omega transformation during isothermal annealing of refractory high entropy alloys: Implications for high temperature phase stability

Al containing refractory high entropy alloys or complex concentrated alloys (RHEAs or RCCAs) often exhibit a patterned BCC + B2 microstructure, resembling the FCC + L12 microstructure typically observed in many Ni/Co base superalloys. There is a rapidly growing interest in these BCC+B2 microstructures, especially due to their promising balance of elevated temperature mechanical properties. Unfortunately, the stability of such microstructures on long-term exposure to elevated temperatures is not well-understood and prompts the motivation of the present study. The current study traces the phase transformation pathway during the intermediate temperature annealing of such microstructures for extended time periods, and more specifically the formation of ordered omega type phases. In the case of a high strength RHEA, Al0.5Mo0.5NbTa0.5TiZr the relationships between the two structures strongly suggest a strong possibility that such ordered omega phases are essentially derived from a parent B2 phase. At the early stages of annealing, the solutionized condition of this alloy consists of a co-continuous mixture of BCC and B2 phases which transforms into a continuous B2 matrix with discrete BCC precipitates presumably via spinodal decomposition concomitant with ordering. With increasing annealing time at 800 °C, isolated islands of BCC phase embedded with discrete, fine-scale pockets of the B2 and hP18 (P63/mcm) phase develop, where the hP18 phase is an ordered derivative of the omega phase, commonly found in titanium and zirconium base alloys. The transformation of the parent B2 phase to the hP18 ordered omega type phase can be rationalized via a combined replacive and displacive transformation, including simultaneous rejection and enrichment of specific alloying elements to achieve the desired stoichuometry. First-principles based DFT calculations were also carried out to compare the stability of the B2 versus hP18 phases, as a function of composition. While long term annealing and CALPHAD based calculations both indicate towards an equilibrium phase field comprising of hP18, B2 and BCC phases, hP18 does not directly nucleate homogeneously within the BCC grains. Instead, it appears to form heterogeneously along grain boundaries and more importantly within the grains with B2 phase acting as an intermediate phase, both of which indicates a high nucleation barrier for its homogeneous nucleation. The universality of this phenomenon has been demonstrated using examples from other RHEAs, and the resultant impact on the high temperature stability of the BCC+B2 microstructure has been discussed.

Suppressed hydride precipitation in the welding zone of a zirconium-based alloy cladding tube

The difficulty in characterizing mesoscale hydrides in the highly deformed submicron (< 1 μm) scale grain size of the reactor-grade stress-relief annealed (SRA) zirconium-based alloy tube via electron backscatter diffraction (EBSD) prevented, so far, the field from directly observing the mesoscale characteristics of precipitated hydrides in the welding and tube region. This study presents new high-quality EBSD characterization of precipitated hydrides along the axial length of reactor-grade zirconium-based alloy tubing covering the end cap head, welding zone (WZ), heat affected zone (HAZ), and nominal tube region. The result demonstrates that hydride precipitation is suppressed in the WZ, indicating increased resistance of the hydride embrittlement. Extensive EBSD analyses coupled to thermodynamic modeling were conducted to unveil the mechanism of hydride precipitation suppression at the WZ. Suppressed hydride precipitation in the WZ region is caused primarily by the combined effects of a higher critical nucleation Gibbs free energy of the WZ (ΔGWZ*) due to a greater matrix stiffness, increased misfit strain, and greater grain boundary misorientation angles.

Research on Chemical Mechanical Polishing Technology for Zirconium-Based Amorphous Alloys

Crystallization often occurs in the processing of amorphous alloys, causing the materials lose their excellent properties. The study adopts chemical mechanical polishing of amorphous alloys, presenting the effect of the rotational speed of the polishing turntable, size of abrasive, polishing pressure, and oxidant concentration. The Taguchi method is used to find the best processing parameters, and AFM is used to characterize the machined material surface. At the same time, XPS is used to detect the change of oxide film composition with the addition of oxidant. The results indicate the optimum process parameters: rotational speed of the polishing turntable is 75 r/min, polishing pressure is 28.3 kPa, the size of abrasive is 0.5 μm, and the size of abrasive is a significant factor affecting surface roughness Sa. In addition, as the size of abrasive increases, the material removal rate increases while the surface roughness Sa increases. At pH 10, with an abrasive particle size of 0.5 μm, as the H2O2 concentration increases, the MRR first rapidly decreases at 0.21 wt.% H2O2, and then gradually increases, while the Sa decreases. Furthermore, with the addition of oxidant, the main composition of the surface oxide film changes from oxide to hydroxide, and the contents of Zr4+ and Cu0/Cu1+ elements increase. The findings can provide a feasible chemical mechanical polishing process for zirconium-based amorphous alloys to obtain a satisfactory polishing effect.

Evolutionary design of machine-learning-predicted bulk metallic glasses.

The size of composition space means even coarse grid-based searches for interesting alloys are infeasible unless heavily constrained, which requires prior knowledge and reduces the possibility of making novel discoveries. Genetic algorithms provide a practical alternative to brute-force searching, by rapidly homing in on fruitful regions and discarding others. Here, we apply the genetic operators of competition, recombination, and mutation to a population of trial alloy compositions, with the goal of evolving towards candidates with excellent glass-forming ability, as predicted by an ensemble neural-network model. Optimization focuses on the maximum casting diameter of a fully glassy rod, D max, the width of the supercooled region, ΔT x, and the price-per-kilogramme, to identify commercially viable novel glass-formers. The genetic algorithm is also applied with specific constraints, to identify novel aluminium-based and copper-zirconium-based glass-forming alloys, and to optimize existing zirconium-based alloys.