Zircaloy Research Articles

Laser cutting study of zirconium alloys for nuclear decommissioning

We conducted laser cutting studies on zirconium alloys, specifically Zircaloy-2 and Zr-2.5%Nb alloy, which are used as constituent materials in the nuclear fuel channel of a pressurized heavy water reactor. The study measured the maximum cutting speed, amount of secondary emissions, and aerosol characteristics for each material using 10 mm thick plate specimens. The cutting performance of the zirconium alloys was similar to each other. At a laser power of 1–5 kW, the maximum cutting speed ranged from 750 to 1900 mm/min, and the amount of secondary emissions per length ranged from 32 to 53 g/m. Compared to 304 L stainless steel, the maximum cutting speed was 1.7–1.9 times higher, and the amount of secondary emissions was about 60–70 % of that of stainless steel. Analyzing the physical properties of aerosols, both zirconium alloys generated larger particles with a count median aerodynamic diameter of 0.25 μm, which is approximately 15–17 % larger than that of 304 L stainless steel.

Graphene protective coatings for hydrogen resistance improving of E110 zirconium alloy

This work is devoted to the study of barrier properties of graphene coating against hydrogen permeation into E110 zirconium alloy. For the first time, the kinetics of hydrogen absorption by the graphene-coated zirconium alloy with a single-layer graphene coating obtained by chemical vapour deposition (CVD) is studied in this work. It was shown that the graphene coating exhibits barrier properties and reduces hydrogen absorption rate by E110 zirconium alloy in the temperature range (350–550)°C. The activation energy for hydrogen absorption was reduced from 72 to 61 kJ/mol. Phase transformations in zirconium alloy during hydrogenation can result in degradation of protective properties of graphene coating.

An Integrated Solution to FIB-Induced Hydride Artifacts in Pure Zirconium.

The preparation method of transmission electron microscopy (TEM) samples for pure zirconium was successfully executed using a focused ion beam (FIB) system. These samples unveiled artifact hydrides induced during the FIB sample preparation process, which resulted from stress damage, ion implantation, and ion irradiation. An innovative solution was proposed to effectively reduce the effect of artifact hydrides for FIB-prepared samples of hydrogen-sensitive materials, such as zirconium alloys. This development lays the groundwork for further research on the micro/nanostructures of zirconium alloys after ion irradiation, thereby facilitating the study of corrosion mechanisms and the prediction of service life for nuclear fuel cladding materials. Furthermore, the solution proposed in this study is also applicable to TEM sample preparation using FIB for other hydrogen-sensitive materials such as titanium, magnesium, and palladium.

Improvement and prediction technology of the water-side corrosion of zirconium alloy: the developmental tendency

Abstract The cladding material serves as the primary safety barrier in nuclear power plants, and its performance significantly impacts the service life and safety reliability of fuel elements. Zirconium-based alloys (Zr), widely utilized as cladding materials in light water reactors (boiling water reactors and pressurized water reactors), face numerous challenges in an environment characterized by increasing fuel consumption, with corrosion being the foremost concern. The corrosion of Zr alloys involves multiple influencing factors, including temperature, chemical environment of water, evolution process of second phases, destruction-regeneration cycle of oxidation films, stress accumulation, volume changes, defect formation, and irradiation effects. This paper provides a comprehensive review on the field of aqueous-side corrosion of Zr alloys while also discussing technical approaches to enhance their corrosion resistance properties and presenting current development status regarding predictive technologies for assessing corrosion behavior along with predicting future trends.

The role of symmetric tilt grain boundaries on the precipitation of hydrides in zirconium: A molecular dynamics study

Zirconium (Zr) alloys are widely used as structural materials in the core of nuclear reactors. However, the absorption of hydrogen by these alloys during the reactor operation may lead to the precipitation of hydrides. Such hydrides have considerable effects on the microstructure and mechanical properties of the alloys, thus significantly impacting their overall performance. In this work, we investigate the behavior of hydrogen in the presence of symmetric tilt grain boundaries in Zr using molecular dynamics simulations. To systematically explore the conditions for the formation of hydrides, we consider hydrogen concentration levels of up to 50 at.%. We find that hydrogen concentrations of 30 at.% or lower are below the threshold for the precipitation of hydrides, while fcc coordinated hydrides precipitate at 50 at.%, which is very close to the solubility limit of hydrogen in Zr at 800 K. We find that the grain boundaries (GBs) play a pivotal role in the behavior of hydrogen, as well as the formation of hydrides in Zr systems. In most cases, large amounts of hydrogen are found to accumulate at GBs. However, both homogeneous and heterogeneous nucleation are observed during the formation of hydrides at the grain interiors and GBs, respectively. Nonetheless, in some cases the GB inhibit the formation of hydrides.

Effect of pulse laser nitriding on the superheated steam corrosion behavior of Zr alloy

Zirconium (Zr) alloy fuel cladding is exposed to harsh environments such as high temperature and high pressure for a long time. Surface modification technology can effectively improve its high-temperature oxidation resistance. In this work, pulse laser nitriding technology was used to perform laser nitriding treatment on Zr alloys at different laser energies. The effect of different energy levels of laser nitriding treatments on the superheated steam corrosion resistance of Zr alloys was studied. First-principles was used to reveal the mechanism of how the diffusion layer improves the performance of Zr alloys in resisting superheated steam corrosion after nitriding treatment. Results indicate that the surface of Zr alloys treated with laser nitriding exhibits a molten feature. When the laser energy is 100 mJ, N mainly exists in the form of a diffusion layer in the Zr alloy substrate. As the laser energy rises to 200 mJ and above, the ZrN phase is generated. After 1000 h of superheated steam corrosion, the corrosion weight gain of the 100 mJ sample was reduced by approximately 50 % compared to the Zr alloy. The corrosion weight gain of the 400 mJ sample increased by approximately 12 %. The presence of the ZrN phase accelerates the oxidation degradation rate of Zr alloy. When N atoms exist as a diffusion layer, the transition from tetragonal zirconia to monoclinic zirconia can be suppressed. First-principles calculations showed that when O and N atoms coexist in the Zr alloy substrate, their binding performance with surface O is significantly decreases, thus further improving the 100 mJ sample's resistance to superheated steam corrosion.

A Comparative Study on Force-Fields for Interstitial Diffusion in α-Zr and Zr Alloys.

Interstitial diffusion is important for radiation defect evolution in zirconium alloys. This study employed molecular dynamics simulations to investigate interstitial diffusion in α-Zr and its alloys with 1.0 at.% Nb and 1.0 at.% Sn using a variety of interatomic potentials. Pronounced differences in diffusion anisotropy were observed in pure Zr among the employed potentials. This was attributed to the considerable differences in migration barriers among the various interstitial configurations. The introduction of small concentrations of Nb and Sn solute atoms was found to significantly influence diffusion anisotropy by either directly participating in the diffusion process or altering the chemical environment around the diffusing species. Based on the moderate agreement of interstitial energetics in pure Zr, accurately describing interstitial diffusion in Zr alloys is expected to be more complex. This work underscores the importance of the careful validation and selection of interatomic potentials and highlights the need to understand the effects of solute atoms on interstitial diffusion.

Phase evolution of Al–Al2O3–ZrO2 refractories at 1300 °C under N2 flowing

In order to obtain Al–Al2O3–ZrO2 composite refractories with excellent oxidation resistance, erosion resistance and thermal shock stability, metallic aluminium and fused Al2O3–ZrO2 were used as the raw materials, while phenolic resin was used as the binding agent. The microstructure was characterised by means of X-ray diffraction and scanning electron microscopy tests. Based on the reaction between zirconia and metallic aluminium under N2 flowing at 1300 °C, the reaction mechanism of zirconia and metallic aluminium in-situ reaction at 1300 °C to generate (Al2OC) x(AlN)1− x solid solution, zirconium nitride, and aluminium–zirconium alloy was revealed, and a model for the evolution of the phases in the aluminium-added Al–Al2O3–ZrO2 specimen was constructed to realise the Al–Al2O3–ZrO2 composite refractories phase controlling. Based on the good corrosion resistance of ZrO2 to CaO and the excellent results of the in-situ reaction of metallic aluminium to generate a large number of non-oxide reinforcing phases capable of replacing elemental C, this work is of great significance for the study of refractory in erosion-heavy environments, such as calcium-treated steel.

Effect of Mo Content on the Structural, Mechanical, and Tribological Properties of New Zr-Nb-Mo Alloys Obtained by Combining Powder Metallurgy and Vacuum Arc Melting Methods.

Considering the high demand for innovative solutions in medicine, a major increase in interest in biomaterials research has been noticed, with the most significant advancements in metals and their alloys. Titanium-based alloys are one of the most recognised in the scientific community but do not represent the only way to achieve optimal results. Zirconium alloys for medical applications are a novelty with significant research potential based on their outstanding properties, which may be of value for medicine. The aim of the present study was to obtain new biomedical Zr-Nb-Mo alloys with varying ratios of their respective elements-Zr and Mo-using combined powder metallurgy (PM) and arc melting (VAM) methods. The obtained samples underwent microstructure analysis using an optical microscope (OM) and a scanning electron microscope (SEM). The study of element distribution was conducted with energy dispersive spectroscopy (EDS), whereas the phase composition was determined using X-ray diffraction (XRD). Mechanical properties were examined with a Micro Combi Tester MCT3, whereas tribological properties were assessed with a TRN Tribometer, and Ringer's solution was used as a lubricant. Additionally, the wear tracks of the studied samples were observed using the SEM. The research results indicated that increased Mo content conduced to microstructure refinement and homogeneity. Furthermore, the higher content of this element contributed to the growth of the HVIT, HIT, and EIT parameters, together with the improvement in the tribological performance of the alloys. XRD analysis revealed that the obtained samples were multiphase, and raising the Mo addition promoted the formation of new phases, including a ternary phase-Zr0.9Nb0.66Mo1.44 (Fd3¯m). The chemical composition study showed uneven distribution of niobium and areas of uneven mutual distribution of zirconium and molybdenum.

Helium bubble evolution in Zr alloys: Effects of sinks and temperature

Implanted helium resulting from fusion reactions can pose a significant threat to zirconium alloys used as fusion reactor materials. Intrinsic structures and temperatures in nuclear materials can significantly influence the behavior and distribution of these helium bubbles under irradiation. Therefore, the behavior of helium bubbles in zirconium alloys was investigated by in-situ He+ irradiation at 573 K, 623 K, 673 K, 723 K and 773 K. Results indicated that the smallest and densest bubbles are found at GBs, while the largest helium bubbles are observed near precipitates. The size of the helium bubbles first increased and then decreased with an increase in temperature, while the density of bubbles changed in the opposite trend. The swelling induced by irradiation was maximum at 673 K when the irradiation dose was constant, while at this temperature, the geometry of the helium bubble changed from spherical to polygonal with a larger size than those at other temperatures.

Improved understanding of the growth of blocky alpha in welded Zircaloy-4

Zirconium alloys are widely used in nuclear reactors as fuel cladding materials. Fuel cladding is used to contain the nuclear fuel and cladding tubes are typically sealed using welds. Welding of zirconium alloys can result in changes in the local microstructure, with the potential to grow so called ‘blocky α’ grains in the welded region during subsequent thermal processing and these blocky α grains have the potential to be detrimental to the integrity of the component. In this work, complimentary heating experiments with ex situ and in situ electron backscatter diffraction (EBSD) analysis are used to aid understanding of the blocky α grain growth within the weld region. These experiments reveal that blocky α grain growth is related to the prior-β (high temperature) microstructure, as grains grow adjacent to a prior-β grain boundary and the orientation of the growing α grain can be explained using neighbourhood orientations from this prior-β grain. This growth mechanism is explained via a simple mechanism which is related to the α grain orientations, grain boundary structures and local stored energy. Ultimately, our findings indicate that the likely grain growth (size and morphology) across the weld region can now be predicted from the initial as-welded microstructure.

The micromechanics of fracture of zirconium hydrides

Zirconium alloys are susceptible to hydrogen embrittlement and hydride precipitation. The precipitation of hydrides is accompanied by a transformation strain, but the contribution of this strain to the fracture of hydrides is not well-understood. In this study, deformation mechanisms and the micromechanics of fracture of hydrides are investigated by conducting in-situ and interrupted ex-situ tensile experiments on hydrided zirconium specimens. Electron backscatter diffraction (EBSD) is used to measure the macro-EBSD maps that contain the grains located in the gauge section of the specimens’ surfaces. Further, high spatial resolution EBSD is used to determine orientation variations induced by the precipitation of hydrides and by the external mechanical load. The measured grain orientations are mapped into a crystal plasticity finite element (CPFE) model to examine the performance of eight different crack initiation criteria. It is shown that a multiscale approach is essential for studying the fracture of hydrides as the details of hydride morphology, orientation, local grain neighborhood, and hydride-hydride interactions are important in such analysis. It is shown that neglecting the effects of hydride-induced transformation strain leads to inaccuracies in predicting both the location and direction of microcracks within hydrides. Among the examined methods, the combination of resolved shear stress and resolved shear strain on slip systems, i.e., the highest shear energy density, consistently predicts the correct locations of hydride microcracks as well as their propagation direction. Further, it is shown that the significant deformation that takes place within hydrides is the main driving force for the fracture of hydrides.

Effect of Discharge Energy on Micro-Arc Oxidation Coating of Zirconium Alloy.

The micro-arc oxidation (MAO) technique was used to grow in situ oxidation coating on the surface of R60705 zirconium alloy in Na2SiO3, Na2EDTA, and NaOH electrolytes. The thickness, surface morphology, cross-section morphology, wear resistance, composition, and structure of the micro-arc oxidation coating were analyzed by an eddy current thickness measuring instrument, XPS, XRD, scanning electron microscopy, energy spectrometer, and wear testing machine. The corrosion resistance of the coating was characterized by a polarization curve and electrochemical impedance spectroscopy (EIS). The results show that, with the increase in frequency, the single-pulse discharge energy decreases continuously, and the coating thickness shows a decreasing trend, from the highest value of 152 μm at 400 Hz to the lowest value of 87.5 μm at 1000 Hz. The discharge pore size on the surface of the coating gradually decreases, and the wear resistance and corrosion resistance of the coating first increase and then decrease. The corrosion resistance is the best when the frequency is 400 Hz. At this time, the corrosion potential is -0.215 V, and the corrosion current density is 2.546 × 10-8 A·cm-2. The micro-arc oxidation coating of zirconium alloy is mainly composed of monoclinic zirconia (m-ZrO2) and tetragonal zirconia (t-ZrO2), in which the content of monoclinic zirconia is significantly more than that of tetragonal zirconia.

Redistribution of Nb and other alloying elements in Nb-doped Zr alloy under high dose ion irradiation

ABSTRACT To understand the microstructural changes in Nb-doped zirconium alloy cladding at high irradiation doses, the Mitsubishi Developed Alloy (MDA) cladding was ion- irradiated up to 40 dpa at 400 ◦ C and the distribution of alloying elements was in- vestigated using atom probe tomography (APT) and scanning transmission electron microscopy and energy dispersive X-ray spectroscopy (STEM-EDS). APT analysis revealed two types of Nb nanoclusters with different Nb contents after the ion irradia- tion. The Nb concentration in the matrix decreased significantly with ion irradiation up to 10 dpa and then decreased slowly to 40 dpa. STEM-EDS analysis estimated that the secondary phase precipitates (SPPs) of Zr(Fe,Cr,Nb)2 and Zr(Fe,Cr)2 were formed before the ion irradiation. Although Fe, Cr, and Nb were dissolved from the SPPs during the ion irradiation, these alloying elements still remained in the SPPs after 40 dpa irradiation. Other Fe-Cr nanoclusters, Nb-rich phase, and Fe-rich phase were also formed during the ion irradiation. These radiation-induced precip- itates were observed until after 40 dpa irradiation and were expected to be stable under irradiation conditions.

Effect of grain boundary character on intergranular hydrides precipitation in zirconium

Grain boundary characteristics play a crucial role in the formation of intergranular hydrides, which may cause brittle fracture in zirconium alloys. However, the impact of grain boundary character on microscopic hydride precipitation remains unclear. Here, we investigate the nucleation and precipitation of intergranular hydrides in zirconium by using In-situ transmission electron microscopy observations and post-precipitation electron backscattered diffraction analysis. The nucleation of intergranular hydrides is highly dependent on the interacting angle (αGB-BP) between the grain boundary plane and the basal plane. For low-angle grain boundaries, hydrides prefer to grow into the grain interior when the αGB-BP is less than 60°, otherwise, no hydride precipitation occurs. For high-angle grain boundaries, hydrides nucleate along the grain boundary when the αGB-BP is less than 39°. As the αGB-BP increases, hydrides grow into the grain interior. A thermodynamic model is developed to analyze the nucleation of needle-shaped hydrides. The variation of the intergranular hydride as a function of the grain boundary energy and the αGB-BP is predicted. This finding provides new insight into utilizing grain boundary engineering to modulate intergranular hydride precipitation and offers a pathway to control hydrogen embrittlement in zirconium alloys.

The effect of chromium-based coatings on corrosion behavior of alloy Zr1Nb in 70ppm Li+ water environment

Recent research indicates that one of the leading candidates for Accident Tolerant Fuels (ATF) are chromium-based coatings deposited on the commercially used zirconium alloys. The chromium-based coatings seem to improve the corrosion kinetics of underlying zirconium in both primary water and steam, where zirconium fails to meet the safety requirements. It is well known that the corrosion kinetics of zirconium can be negatively influenced by high concentrations of lithium ions. To evaluate the effect of chromium-based coatings on corrosion behavior of zirconium alloys in water containing elevated concentrations of lithium, Cr and CrN/Cr multilayer coated Zr1Nb alloy was tested in high temperature water containing 70 ppm Li+. Using Electrochemical Impedance Spectroscopy (EIS), Scanning Electron Microscopy (SEM) and Energy-dispersive X-ray Spectroscopy (EDX), we have found that the chromium coating could have a positive effect on the underlying material as it behaves as a good diffusion barrier for oxygen and Li+ions. On the other hand, CrN/Cr coating, due to not sufficient structural integrity of the multilayer, showed non-protective behavior to the Zr1Nb alloy. Our results suggest that certain chromium coatings can significantly enhance corrosion resistance in lithium containing water environments.

Resarch on Microstructure and Property of Zirconium alloy in the Rotary forging process by GNN-FEM

Resarch on Microstructure and Property of Zirconium alloy in the Rotary forging process by GNN-FEM

Effects of nitrogen content on the phase composition and high-temperature steam oxidation resistance of Cr0.5Al0.5Nx (x = 0.33, 0.81, 1.23) coatings on Zry-4 alloy

Cr0.5Al0.5Nx (x = 0.33, 0.81, 1.23) coatings were deposited on Zry-4 alloy using magnetron reaction sputtering technique to investigate their application as accident-tolerant fuel materials. The effects of nitrogen (N) content on the phase composition and high-temperature steam oxidation resistance of the coatings were systematically investigated. With the increasing N content, the phase composition of the coatings is mainly transformed from Cr(Al) to the cubic-CrAlN phase, which may be due to the formation of a greater number of CrN bonds and AlN bonds. The high-N content of the Cr0.5Al0.5Nx coating can significantly enhance the high-temperature steam oxidation resistance of zirconium alloy. Compared to the bare Zr alloy, the weight gain oxidation of the Cr0.5Al0.5N1.23 coating is reduced by 45 % after oxidation at 1200 °C for 1 h. The oxide layer thickness of the bare Zry-4 alloy is as high as 350 μm, while the oxide layer thickness of the Cr0.5Al0.5N1.23 coating is only about 12 μm. The complex oxide layer structure and an internal continuous Al2O3 layer are formed, effectively preventing the inward diffusion of O2−. In addition, with the increase of N content in the coatings, the degree of interdiffusion between the coating and the substrate is weakened. The formation of α-Zr(N) layer in the Cr0.5Al0.5N1.23 coating significantly inhibits the interdiffusion between the coating and the substrate.

On corrosion of zirconium alloy in dissolved oxygen water: The role of Cu addition

It is significantly important to improve the corrosion resistance of Zr-Sn-Nb alloys in dissolved oxygen (DO) high temperature water by compositional modification. Here, the effect of Cu addition on the corrosion behaviors of Zr-0.5Sn-0.2Nb-0.4Fe-0.2Cr alloy was studied by using an autoclave circulating water loop at 360 °C/20 MPa up to 250 days. The oxide microstructures were characterized by a combination of SEM, TEM, STEM-EELS and PED. The results showed that compared with 0Cu alloy and 0.07Cu alloy, 0.13Cu alloy shows better corrosion resistance and its final weight gain after 250 days exposure is even lower than that of commercial Zr-4 alloy. The oxides for 0.07Cu and 0.13Cu alloys show more ordered crystals, thinner oxide and less lateral cracks, and the columnar grains show a stronger {10–3} texture, compared to Cu-free counterpart. The t-Zr2Cu SPPs have oxidized in the oxygen-rich layer below the O/M interface. The Cu ions tend to diffuse towards the grain boundary of dense columnar grains and maybe strengthen the grain boundaries, avoiding the defect generation under intrinsic stress in the oxide.

Protective layers of zirconium alloys used for claddings to improve the corrosion resistance

Abstract Zirconium alloys are used as a cladding material for fuel elements in nuclear reactors. In the case of severe accident conditions, the possible rapid oxidation of zirconium in steam or/and air may result in intense hydrogen generation and hydrogen–oxide mixture explosion. Advanced technologies for increasing the corrosion resistance of claddings are being investigated in two directions: (a) protective coatings on Zr alloys and (b) the use of new materials for claddings. Coatings with silicon may provide a more protective barrier than the ZrO2 films formed on an alloy cladding during nuclear plant operations. These coatings may also serve as a protective barrier during high-temperature accidents. The current work aimed at developing protective coatings with silicon on zirconium alloys. Multielemental Zr–Si–Cr coatings were formed on Zry-2 alloy using the physical vapor deposition (PVD) method. Long-term oxidation tests were carried out under the following conditions: 360°C/195 bar/63 days/water-simulating PWR water. Obtained results show the protective character of formed layers. The material in the form of silicon carbide grains covered with yttrium–aluminum garnet (SiC + YAG) was prepared using the sol–gel method. The formed powder is the main component for coating formation on Zr–1Nb alloy using the method of suspension plasma spraying (SPS).