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The accuracy and reliability of small-scale mechanical tests remain doubtful due to significant dependence of the obtained mechanical properties on specimen size. Mesoscale tensile tests with specimen sizes ranging from 10 μm to 1 mm are capable of obtaining bulk-like properties but are rarely applied to hexagonal close-packed metals. In this study, well-designed comparative tensile tests were carried out on a Zircaloy-4 alloy with a grain size of 4 μm using femtosecond laser-machined mesoscale specimens with a thickness of about 60 μm, sub-sized specimens with a thickness of about 1.3 mm, and standard specimens with a thickness of 4 mm. The quantitative results revealed that irrespective of the small specimen dimensions, the yield strength, tensile strength, and tensile ductility are only approximately 10.4%, 5.2%, and 13% lower than those of the standard specimens, respectively. This clearly demonstrates that the mechanical properties can be assessed with satisfactory accuracy by mesoscale tensile tests. The comparatively greater deviation of the yield strength at the mesoscale arises from the disappearance of yield point behavior, while the reduced tensile ductility is associated with the larger volume fraction of surface grains. The surface grains are characterized by more surface dislocation sources and deform with weaker constraints from neighboring grains, leading to smooth plastic yielding and slightly reduced strain hardening at the mesoscale. [ABSTRACT FROM AUTHOR]

A numerical model for hydride embrittlement in Zirconium alloy (Zr–2.5Nb) is developed utilizing the extended finite element method (XFEM). Hydride embrittlement reduces the ductility and failure time of a metal/alloy. During hydride embrittlement, stress-directed hydrogen diffusion, metal-hydride phase transformation, mechanical deformation, and hydride precipitation occur simultaneously. The present model incorporates all these processes and is able to predict the hydrogen concentration and the hydride fraction distribution under any externally applied stress field. In this work, both the steady and transient hydrogen diffusion cases are evaluated. Further, the XFEM is utilized to develop a model of hydride embrittlement in the presence of a crack. The first step of the hydride embrittlement process is the diffusion of hydrogen. According to Fick's law of diffusion, hydrogen diffusion is directly dependent on hydrostatic stresses and hydrogen concentration gradient under external stresses. The next step is the hydride precipitation in hydride embrittlement, where the expansion of material takes place that changes the hydrostatic stress field. Thus, studying the effect of precipitation of hydride on hydrostatic stresses is essential. Moreover, the process of hydride embrittlement is highly influenced by residual stresses in the structure. Hence, the effect of residual stress present in the zirconium alloy pressure tube (PT) is also evaluated. The results indicate that the residual tensile stresses contribute to the growth of hydride, which will reduce the material failure time. [ABSTRACT FROM AUTHOR]

Hydrogen embrittlement is a critical issue for zirconium alloys, which receives long-term attention in their applications. The formation of brittle hydrides facilitates crack initiation and propagation, thereby significantly reducing the material's ductility. This study investigates the tensile properties and hydride morphology of a novel zirconium alloy under different hydrogen-charging current densities ranging from 0 to 300 mA/cm2, aiming to clarify the influence of hydrides on the fracture behavior of the alloy. The mechanical property results reveal that, as the hydrogen-charging current density increases from 0 to 100 mA/cm2, the maximal elongation decreases from 24.99% to 21.87%. When the current density is further increased from 100 mA/cm2 to 200 mA/cm2, the maximal elongation remains basically unchanged. However, a substantial drop in elongation is observed as the hydrogen-charging current density rises from 200 mA/cm2 to 300 mA/cm2, decreasing from 20.77% to 15.18%, which indicates a marked deterioration in hydrogen embrittlement resistance. Subsequently, phase compositions, fracture morphology, and hydride types in the fracture region of tensile specimens were characterized. The morphology and quantity of hydrides change with increasing hydrogen-charging current density. When the hydrogen-charging current density reaches 100 mA/cm2, the δ-phase hydrides form, which significantly reduces the ductility of the zirconium alloy. At a hydrogen-charging current density of 200 mA/cm2, metastable ζ-phase hydrides are formed, resulting in negligible variations in the alloy's mechanical properties. However, when it comes to 300 mA/cm2, stable δ-phase hydrides with diverse morphologies form, leading to a pronounced degradation in tensile performance. Finally, by integrating mechanical tests with microstructural characterization, the influence of hydrides formed under different hydrogen-charging current densities on the zirconium alloy was analyzed. With increasing hydrogen-charging current density, the maximal elongation of the specimens gradually decreases, while the tensile strength steadily increases. At a hydrogen-charging current density of 300 mA/cm2, a larger amount of hydrides is formed, and the hydride type transitions completely from a mixture of δ-phase and ζ-phase hydrides to entirely δ-phase hydrides. The formation of lath-like δ-phase hydrides induces twinning structures, resulting in further lattice mismatch, which significantly reduces the maximal elongation of the zirconium alloy. Additionally, the morphology of the δ-phase hydrides changes from slender needle-like structures to lath-like structures, leading to a notable increase in internal stress, which in turn further enhances the tensile strength of the specimens. [ABSTRACT FROM AUTHOR]

The study focuses on developing a robust method for determining zirconium in U–Pu–Zr alloy samples, crucial for assessing fuel composition and ensuring uniformity. Using mandelic acid precipitation and spectrophotometric quantification with Arsenazo-III, the method achieves high sensitivity and reproducibility, crucial for handling microgram-levels of Pu. By optimizing conditions to maintain Pu in its + 3 oxidation state, interference during Zr precipitation is minimized. Comparative analyses with ICP-AES validate the accuracy and reliability of the results. This approach not only enhances analytical precision but also reduces radioactive waste, making it suitable for nuclear fuel characterization. [ABSTRACT FROM AUTHOR]

Zirconium alloys used in nuclear reactor exhibit, under fast neutron flux, a macroscopic deformation even without applied stress called irradiation induced growth. Because of the polycrystalline nature of the material, the local growth of individual grains results in strain incompatibilities yielding to intergranular stresses and thus to local creep. In order to study and understand the influence of the local anisotropic creep on the macroscopic growth behaviour, an analytical and a numerical study have been undertaken, using Voigt and self-consistent estimates and also fast Fourier transform simulations. It is shown that the anisotropic local creep has a strong influence on the effective macroscopic growth strain of the polycrystal. Especially, when the deformation is difficult along the $\langle c\rangle$ 〈 c 〉 axis, a growth enhancement effect is observed. This phenomenon is well explained in the frame of the Voigt estimate using a reduced fibre texture. Computations conducted using a texture representative of the industrial material provide a quantitative confirmation of this enhancement effect. This work demonstrates the significant contribution of the local anisotropic creep to the macroscopic in-reactor growth strain of zirconium alloys. [ABSTRACT FROM AUTHOR]

In this paper, we report our findings on the preparation of a film/alloy/substrate system with the use of a low-energy high-current electron beam (LEHCEB), as a result of sequential growth of molybdenum and chromium films on a zirconium substrate by magnetron sputtering. The system was produced in a single vacuum cycle, by cyclic growth of molybdenum film on a substrate and LEHCEB irradiation, followed by chromium film growth as a final step. We demonstrate that LEHCEB irradiation leads to the formation of Mo–Zr alloy consisting mainly of the high-temperature phase β-Zr. The alloy has the form of a layer 4.2 ± 0.9 μm in thickness, containing molybdenum-enriched regions. Such regions result from the formation of secondary phases: intermetallic compound Mo2Zr and a solid solution of Zr in Mo. [ABSTRACT FROM AUTHOR]

The robust performance of components within nuclear fuel reprocessing plants, especially in environments characterized by hostility, corrosion, and radioactivity, is imperative for ensuring seamless plant operations. Nitric acid serves as the primary process medium in the complex chemical processes involved in reprocessing spent fuel. While 304L SS is conventionally employed for nitric acid service in these plants, there is an ongoing exploration of alternative candidates such as titanium (Ti), zirconium (Zr) and their alloys. This study focuses on assessing the performance of commercially pure titanium (CP-Ti) and zirconium alloy (Zr-4) in nitric acid with formaldehyde at 80 °C. The corrosion rate of the Zr-4 sample, immersed in nitric acid with formaldehyde for two weeks at 80 °C, was negligible when compared to the commercially pure titanium (CP-Ti) sample under the same conditions. Comprehensive characterization of the oxide film formed on the surface of CP-Ti and Zr-4 samples was achieved through field emission scanning electron microscopy (FE-SEM) coupled with energy dispersive x-ray spectroscopy (EDS), as well as x-ray photoelectron spectroscopy (XPS). The corrosion behavior of both Ti and Zr-4 samples in nitric acid containing formaldehyde was systematically evaluated using potentiodynamic polarization (PDP) and electrochemical impedance spectroscopy (EIS). The results revealed enhanced corrosion resistance for both CP-Ti and Zr-4 samples. Consequently, this study suggests that titanium and zirconium may be considered as suitable candidates for the process equipments handling hot nitric acid with formaldehyde in the aqueous reprocessing of spent fuel. [ABSTRACT FROM AUTHOR]

Cold pilgering technology is one of the most widely used methods to fabricate the seamless zirconium alloy cladding tubes. In practice, the process parameters, such as the reduction ratio of wall thickness and outer diameter and cross-sectional reduction, have not been enough researched in obtaining reasonable control and enhancement of texture during the forming process of zirconium tube. In this study, a more general index, strain ratio, is used to represent the plastic flow and the deformation degree of pilgered tubes. The relationship between strain ratio and texture evolution as well as the influence of tooling design on the strain ratio are quantitatively explored. Besides, a second-order polynomial relationship between basal pole f-parameter and hydride orientation factor has also been developed to achieve the integrated design for fabricating the Zr-Sn-Nb-Fe cladding tubes with high-dimensional accuracy and tailored properties. [ABSTRACT FROM AUTHOR]

After the Fukushima-Daiichi event, there has been notable progress in developing accident-tolerant fuel (ATF) cladding designs to improve light-water reactor (LWR) safety features. In this regard, the Zirconium alloy (Zircaloy) has significant advantages in enhancing the neutron economy, and its unique beneficial characteristics establish it as a reliable and promising material for ATF claddings in the present era. Depositing protective coatings of different interlayer diffusion metals over Zircaloy-based nuclear fuel claddings is becoming increasingly popular as a near-term solution. This strategy can enhance the ability of cladding to withstand accident scenarios, reduce the oxidation rate at high temperatures, and provide additional benefits during both usual and accidental conditions. Current research efforts have strategically prioritized the development of ATF cladding concepts to improve nuclear fuel safety in normal, transient, and potential accidental situations. Herein, we present a comprehensive review of the current research and development events related to the design, manufacturing techniques, and various performance characteristics of chromium (Cr) coated Zircaloy-cladding ATF materials. These materials can improve both the reactor's economics and protection level. Our discussion focuses on these methods for enhancing safety performance in LWRs. In addition, the report provides a prospective viewpoint on future research accomplishments related to these materials. • The necessity for surface modification of Zircaloy is discussed in detail. • Various types of manufacturing processes for Cr-coated ATF cladding materials are elaborated. • Various fabrication parameters that may cause coating failure are deeply examined. • Different performance parameters for Cr-coated Zircaloy cladding are reviewed. • A prospective viewpoint on future accomplishments of Cr-coated ATF claddings is discussed. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Ceramic is a commonly used material in dentistry for reconstructing missing teeth or their tissues due to its biocompatibility, durability and excellent esthetic properties. Despite these advantages, the ceramic restoration damage remains a significant clinical problem. Its causes can be divided into clinical and laboratory factors. The most known include uneven occlusion, improper preparation, trauma, or parafunctions. This study focuses on characterizing less known laboratory causes of ceramic restoration damage. We reviewed the current literature available in the PubMed and Scopus databases. On the basis of 63 selected studies, 3 basic causes of damage were identified: excessive stresses between the framework and ceramic veneering, poor quality of the connection between the facing layer and the substructure, and defects resulting from the nature of the ceramic material such as defects in the ceramic layer, brittleness and lack of flexibility. The stages of the manufacturing process of various permanent ceramic restorations were presented. By controlling these procedures, we can eliminate the errors, resulting in long-term effective functioning of the ceramic restorations. [ABSTRACT FROM AUTHOR]

The application of zirconium alloy in nuclear reactors is supported by its excellent performance and corrosion resistance, but under the action of high-energy neutrons, the atoms in the lattice will change, resulting in the accumulation of lattice defects and the evolution of the microstructure, which not only affect the strength and shape of the material but also affect the tensile resistance, impact resistance and other mechanical properties of the zirconium alloy. Common industrial processes for the preparation of nuclear-grade zirconium sponges are discussed, the advantages, disadvantages and applications of the main production processes are summarised, and the characteristics of irradiation-induced loops are comprehensively examined, encompassing their formation mechanisms, interactions with dislocations, and the range of potential microscopic defects during irradiation. In addition, the phenomena of irradiation growth and irradiation creep and their relationship with microstructure evolution are also discussed. The effect of oxidative corrosion under irradiation has also been explored. Specifically, irradiation can aggravate the corrosion process of zirconium alloy and lead to the thickening of the corrosion layer, but the specific degree of influence is closely related to the working temperature and irradiation conditions. The comprehensive study of zirconium alloy under irradiation conditions aims to provide a scientific basis for the future design of long-life and high-reliability nuclear reactor materials and support the sustainable development of the nuclear industry. [ABSTRACT FROM AUTHOR]

In this study, the effect of zirconium microsegregation occurring during the casting process on the microstructure and recrystallization of the alloy after the modification of an AlMgSi1Mn alloy used in the automotive industry by adding 0.3% Zr by weight was characterized. After the alloys were cast in industrial sizes by the semi-continuous casting method, their microstructures were examined microscopically and the phases formed in the casting were determined. While the average grain size for the as-cast alloy without zirconium was 496.67± 180 µm, the average grain size for the zirconium-containing alloy was found to be 432.52± 167 µm. Electron probe microanalysis was performed to examine zirconium segregation in the ascast alloys. Then, the alloys were homogenized in accordance with industrial conditions. Samples taken from the homogenized alloys were cold deformed by 70% and annealed at 375°C for 15 hours. After annealing, the recrystallization rate was measured as 99.99% in the alloy without zirconium and 12.32% in the alloy containing zirconium. The effect of zirconium segregation on recrystallization behavior was examined by electron probe microanalysis of these samples. As a result of the analysis, it was found that zirconium segregated towards the aluminum dendrite centers during solidification in the casting, and in the regions where the solubility limit was not exceeded in the interdendritic regions, the deformation structure disappeared in these regions after the subsequent deformation and heat treatments, causing recrystallization. [ABSTRACT FROM AUTHOR]

This study has established the radiation shielding efficacy of zircaloy-2 and zircaloy-4 over a wide spectrum of energy levels. Using the Monte Carlo method, the gamma and neutron transmission factors (TF and nTF) were calculated for various energy levels. Zircaloy-2 demonstrated the highest gamma-ray absorption capacity and the lowest neutron absorption capacity among the investigated alloys. The results indicate that zircaloy-2 and zircaloy-4 have nearly the same neutron transmission characteristics. Although many studies have examined the structure and physical characteristics of these materials, there has been a lack of Monte Carlo simulations to comprehensively investigate the correlation between gamma absorption, neutron absorption parameters, and mechanical qualities. This research aims to examine the ability of zirconium and its zircaloy-2 and zircaloy-4 alloys, which are critical materials used in the nuclear industry, to absorb gamma and neutron radiation over a broad spectrum of frequencies. According to the results, zircaloy-2 has the best ability to absorb secondary gamma rays and the highest level of resistance to them. Despite the minimal disparity in the nTF between the two alloys, simulation results have shown that zircaloy-2 has a higher level of neutron transmittance. These results have the potential to expedite the development of novel materials with enhanced attributes for various applications.

Zirconium cobalt alloy (ZrCo) is a promising candidate for tritium storage in fusion reactor, but the surface poisoning caused by trace impurities greatly limits its application. In this work, the interaction mechanisms between nitrogen and ZrCo surface are theoretically explored at the atomic level for the first time. Specifically, the surface exhibits a strong affinity for N atoms with an adsorption energy around −2.0 eV, indicating N can block hydrogenation active sites on ZrCo surface. Seven distinct molecularly adsorbed states have been identified, with their adsorption energies varying from −0.15 to −1.37 eV, depending on both the adsorption site and molecular orientation. The electronic structures and charge distributions indicate that the electrons transfer from ZrCo surface to the anti-bonding orbitals of N 2 , which becomes readily activated with an elongation bond length. Meanwhile, the good mobility of N 2 adsorbate facilitates its dissociation occurring on the specific sites with the energy barriers below 0.3 eV and highly exothermic reaction energies around −3.0 eV, making its dissociative adsorption both kinetically and thermodynamically favorable. Moreover, the high coverage of nitrogen on ZrCo surface significantly restricts hydrogen adsorption and dissociation. Therefore, our results can provide insight into understanding nitrogen poisoning effect on ZrCo surface. • The ZrCo surface exhibits a more affinity for N atoms than H atoms. • The adsorption energy of N 2 is dependent on surface site and molecular orientation. • The good mobility of N 2 adsorbate facilitates its dissociation on the specific sites. • The high coverage of nitrogen significantly restricts hydrogen adsorption and dissociation. [ABSTRACT FROM AUTHOR]

The parameters of the coolant moving in the channel between the fuel elements in the WWER-1200 core with a design steam content of 0.1 at the outlet are determined. The corrosion of the outer surface of these elements is modeled taking into account the heat and mass exchange between the near-wall layer of the coolant flow (with a thickness of the order of the thickness of the evaporating microlayer under the steam bubble) and its core. It is found that the concentration of ammonia in the near-wall layer of the flow is 0.35 of its concentration in the flow core, and this concentration level is typical of the upper half of the core. A formula is proposed for calculating the thickness of the oxidized layer of the outer surface of the fuel elements taking into account the concentration of oxidizers in the near-wall layer of the coolant flow. An increase in the thickness of the oxidized layer due to the effect of boiling is estimated at 30% compared to the thickness observed upon completion of operation during four annual cycles without boiling. [ABSTRACT FROM AUTHOR]

Magnesium alloys, known for their exceptional lightweight properties, have presented challenges in various applications due to their limited corrosion resistance. In this study, the corrosion resistance of Mg97Zn1Y2 magnesium alloys was enhanced by incorporating Zr elements into the Mg97Zn1Y2 matrix, which is distinguished by long periodic stacking ordered (LPSO) phases. Results show that Mg97Zn1Y2-xwt.% Zr (x=0, 0.1, 0.3, 0.6) alloys containing Zr exhibit reduced hydrogen evolution rates and decreased corrosion levels compared with that without Zr, when immersed in a 3.5wt.% NaCl solution. Addition of 0.3wt.% Zr results in the most significant improvement, with a corrosion rate as low as 2.261 mL·cm-2, representing an 86% reduction from 16.438 mL·cm-2 of the base alloy. Furthermore, alloys with Zr additions demonstrate a more positive corrosion potential and lower corrosion current density than does the matrix alloy (64.92 μA·cm-2). The lowest corrosion current density, 21.61 μA·cm-2, occurs with the addition of 0.3wt.% Zr. The introduction of Zr induces a change in the microstructure of the LPSO phases, increasing the charge transfer resistance within the alloy and thus effectively improving its corrosion resistance. [ABSTRACT FROM AUTHOR]

Metals that are used to reconstruct skeletal structures often interfere with magnetic resonance imaging (MRI) owing to differences in magnetic susceptibility; consequently, metals with lower magnetic susceptibilities need to be developed for use in implant devices. Herein, we investigated the corrosion properties of the Zr-14Nb-5Ta-1Mo alloy, which exhibits low magnetic susceptibility and excellent mechanical properties. The pitting potential of Zr-14Nb-5Ta-1Mo was higher than that of pure Zr. The passive current density of Zr-14Nb-5Ta-1Mo also higher than that of pure Zr, which is ascribable to slow reconstruction of the initial passive film associated with the presence of Nb and Ta. XPS revealed that the passive film is enriched with Nb and Ta. Therefore, while the Zr-14Nb-5Ta-1Mo alloy exhibited a high initial passive current density in simulated body fluid, it formed a stable passive film that suppressed localized corrosion. Zr-14Nb-5Ta-1Mo is therefore a prospective implant-material alloy candidate. [ABSTRACT FROM AUTHOR]

Scanning and transmission electron microscopies are used to study the microstructure and phase composition of the surface layer of Zr–1% Nb alloy, which was subjected to treatment by nanosecond laser pulses. During laser treatment, a thin strengthened surface layer with the fine microstructure is found to form. The strengthening of the surface layer no less than 4 µm thick is proved to be due to the formed twin micropackets consisting of martensite nanolamellas and nano-sized ω-Zr phase. [ABSTRACT FROM AUTHOR]

The work described in this article is aimed at investigating the properties of a group of Ti-Zr-Nb alloys. In modern orthopedics and traumatology, the use of materials for bone implants with a minimum modulus of elasticity is becoming increasingly important. This is due to a number of advantages that allow for better integration of the implants with the bone tissue, including the reduction in the detrimental effect of the load-shielding effect, a better load distribution, and stress distribution, which allows for increasing the life of the implant. It is known that the lowest modulus of elasticity in titanium alloys at normal temperature is achieved by the phase composition consisting of metastable β-phase. It is possible to achieve the desired structure by a combination of alloy composition selection and heat treatment. Quenching of titanium alloys allows for the high-temperature β-phase to be fixed. This paper provides justification of the choice of compositions of the studied alloys by calculation methods. The structure of alloys after melting in a vacuum electric arc furnace in an argon environment was studied. The ingots obtained had a dendritic structure. Homogenizing annealing in a vacuum furnace at 1000 °C for 4 h was used to equalize the composition. The structure of the alloyed sheets after hot rolling and hot rolling and quenching was investigated. The microstructure of the plates in both variants had uniform grains of polyhedral shape. X-ray phase analysis of the plates showed that the content of metastable β-phase was 100% before and after quenching. Microhardness testing of the plates showed no significant effect of quenching. The result of the mechanical properties study showed an increase in the plasticity of the material after quenching, with the tensile plots of the samples after quenching reflecting the area where the reverse phase transition of β'<-> α" occurs. Mechanical studies by cyclic loading showed the presence of a superelasticity effect. The Young's modulus study gave a result of 51 GPa for one of the compositions studied. The combination of properties of the materials under investigation has the potential for promising use as a basis for bone implants. [ABSTRACT FROM AUTHOR]

The paper gives an overview of development and application of biocompatible alloys based on zirconium, titanium and niobium, featuring a low modulus of elasticity. The technology of producing a biocompatible 60Zr-20Ti-20Nb alloy and semi-finished products from it in the form of rods and powders for additive manufacturing was developed, and their structure and mechanical properties were studied. The potential for application of the developed biocompatible 60Zr-20Ti-20Nb alloy for manufacturing medical implants is shown. [ABSTRACT FROM AUTHOR]

Within the context of the long-term development plan of batch loading of accident-tolerant fuel assemblies, an equivalent physical model of lead test assemblies is established based on the requirements of radiation testing and the applicability of analysis software. The quantitative assessment has been conducted on the impacts of parameters including zirconium alloy cladding coating and pellet additives on reactivity of fuel assembly. Concurrently, the fuel management strategy is optimized in alignment with the irradiation test objectives of ATF assemblies, leading to the creation of an evaluation process tailored for ATF assembly loading pattern. Focusing on the irradiation of the lead test assemblies into a nuclear power plant in China, three-cycle nuclear reactor loading pattern are designed, which are taking into account the economic, safety, and irradiation testing goals of the nuclear power plant. Simultaneously, the neutronic characteristics are also evaluated, quantitatively. The results show that the established equivalent physical model can effectively represent the characteristics of the lead test assemblies. The loading patterns designed based on the Optimised fuel management strategy achieve the desired goals in terms of economy, safety and radiation test requirements. The designed loading pattern evaluation process and equivalent model analysis method in this work provide guidance for the subsequent engineering design and application of irradiation testing for lead test assemblies. [ABSTRACT FROM AUTHOR]

The effect of nitrogen on an anticorrosive Cr coating for the Zr alloy cladding of nuclear fuel rods is investigated. It is aimed to reduce the number of typical columnar defects generated in Cr‐based coatings when using the cathodic arc ion plating method through reactive nitrogen inclusion and improve the oxidation resistance. Under a working pressure of 5 Pa and an arc current of 60 A in a chrome target, Cr‐based coatings are fabricated under varying N2/Ar gas flow ratios and substrate biases. The growth structures, crystallography, and corrosion behavior of the coatings are characterized using scanning electron microscopy, X‐ray diffraction, and potentiodynamic polarization tests. The Cr coating exhibits a typical columnar structure with voids, pores, and columnar defects. In comparison, through nitrogen inclusion, when the nitrogen content of the Cr‐based coating is less than 15 at%, the Cr(N) coating shows a "featureless structure" similar to an amorphous structure, resulting in fewer voids and defects. Potentiodynamic polarization tests reveal that the Cr(N) coating, featuring a distorted body‐centered cubic‐Cr structure, exhibits a significantly low corrosion current in the passive region. The protection efficiency of the Cr(N) coating is calculated to be 96.9%, confirming its superior substrate protection capabilities compared with chromium nitride coatings. [ABSTRACT FROM AUTHOR]

Integrated zirconium (Zr) alloy fuel claddings require low-temperature and high-efficiency joining technologies for nuclear applications. However, conventional direct-joining technologies make it challenging to satisfy critical requirements at relatively low temperatures in a short time. As a promising alternative, the electric-current-assisted joining (ECAJ) method was studied to reduce the joining temperature and time required for the direct joining process. ECAJ was performed on a spark plasma sintering apparatus, with the specimens assembled with and without graphite dies in the two models. The microstructural evolution and mechanical properties of Zr/Zr joints were investigated at 600 − 900 °C for 1 s to 30 min under a pressure of 30 MPa using different configurations with or without a graphite die. While the joint quality varied with the joining conditions, no significant phase transformation, abnormal grain growth, or preferential orientation was observed at the interface. The maximum shear strength of 404 ± 78 MPa and hardness of 370.9 ± 49.8 HV were achieved at 600 °C for 10 min using a die-less configuration. The joining mechanism was discussed in terms of the potential thermal and electric effects inside the specimen, with the low-temperature joining mechanism attributed to the combined effect of transient overheating and electric-field-accelerated element self-diffusion at the interface, with the latter playing a leading role. [ABSTRACT FROM AUTHOR]

The paper investigates the evolution of the structure and properties of low-alloy dispersion-hardening alloys based on the Cu–Zr, Cu–Cr, and Cu–Cr–Zr systems under high-rate deformation (~105 s–1) by dynamic channel angular pressing (DCAP) and subsequent annealing (aging) at 200–700°C. The effect of alloying with the microadditives Cr (0.09–0.22%) and Zr (0.04–0.20%) in achieving high hardness of copper with a submicrocrystalline structure obtained by DCAP is studied. The effect of DCAP and subsequent aging on the electrical conductivity of alloys is studied. The sequence of decomposition processes of a copper-based α solid solution with the precipitation of nanoscale particles of the second phases and recrystallization is determined. It is shown that the role of zirconium is due to the precipitation of Cu5Zr phase nanoparticles during DCAP and subsequent annealing on dislocations and subboundaries, their fixation, and reduced mobility, as a result, the process of formation of recrystallization centers slows down, which requires rearrangment (restructuring) of the dislocation structure. [ABSTRACT FROM AUTHOR]

Zirconium alloys are widely used in nuclear water reactors as cladding materials. The cladding materials will absorb hydrogen from high temperature water during the operation of nuclear reactor. In cladding tubes, it has been common sense that circumferential hydrides form without stress, while radial hydrides can form when the hydrides are reoriented under stress loading. In this study, we found that a high heating rate can result in hydride reorientation behavior even without stress. At elevated heating rates, the zirconium alloy clad tube developed a non-uniform strain gradient along the direction of heat conduction. Hydrogen atoms migrate preferentially to areas of elevated stress and precipitate as hydrides that are perpendicular to the direction of tensile stress, resulting in the formation of radial hydrides that appear as "sun spots" macroscopically. Additionally, the high heating rate disrupts the {0001}α∥{111}δ, <11–20>α∥<110>δ orientation relationship between the hydride and the substrate, which potentially facilitates crack propagation. [ABSTRACT FROM AUTHOR]

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Zirconium alloys possess excellent wear resistance, which ensures the durability and longevity of the components, making them widely used in medical and other fields. To enhance the functionality of these materials, it is often necessary to fabricate functional microstructures on their surfaces. Electrochemical machining (ECM) techniques demonstrate excellent machining performance for these metals, particularly in the processing of microstructures on complex curved surfaces. However, ECM often faces challenges due to the fluid nature of the electrolyte, resulting in low machining accuracy and localization. This paper proposes a novel method for fabricating complex patterned microstructures using a maskless electrochemical direct writing technique with a polyacrylamide (PAM) polymer electrolyte. By leveraging the non-Newtonian properties of PAM, this method effectively confines the electrolyte to specific areas, thus addressing the issue of poor localization in traditional ECM and reducing stray corrosion. To elucidate the electrochemical removal mechanism of Zr702 in the presence of PAM, polarization curves, viscosity characteristics, and current efficiency parameters were analyzed. Additionally, an experimental study was conducted using a custom-designed nozzle structure. The results showed that the PAM electrolyte could effectively reduce the E F , positively impacting machining accuracy and localization. By controlling the nozzle's motion trajectory, complex microstructures were successfully fabricated through direct writing, demonstrating promising application prospects. [ABSTRACT FROM AUTHOR]

The paper presents the results of the study of the processes of self-propagating high-temperature synthesis of Zr-1%B alloy and its refining by electron beam melting. Experiments on the influence of boron's amorphous and crystalline modifications on the safety parameters of the synthesis process of Zr-1%B alloy necessitated the conversion of amorphous boron into crystalline form by electron beam melting, with an increase in its purity from 94% to 99.9%. High efficiency of vacuum induction and electron beam equipment was demonstrated, which provided a high purity of the Zr-1%B alloy of at least 99.9%. The alloy ingots had a uniform distribution of the alloying element (boron) all over the volume. The obtained alloy is suitable for the production of materials with thermal neutron capture cross-sections of up to 40 barns for neutron protection. [ABSTRACT FROM AUTHOR]

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The anodization or thermal treatments applied to alloys of titanium and zirconium have a substantiated effect on the mixed-oxide layer formation compared to the naturally occurring one. A Ti–Zr 50%/50% alloy was chosen for a comparative study. Controlled, thermally treated, and anodized samples obtained with controlled procedures were analyzed in terms of morphological and compositional analysis (using SEM and EDX analysis) as well as for the determination of hardness variations. Substantial differences were observed depending on the applied functionalization method (compact of structured mixed-oxide nanotubes when the samples are subjected to the anodization procedure); there was an increase of more than six folds in the mixed-oxide layer hardness and D Shore scale, when subjected to thermal treatment, and hence, this lead to the conclusion that one may control the morphology, composition and/or the hardness of the mixed-oxide layer by applying one or another or a combination of functionalization methods. [ABSTRACT FROM AUTHOR]

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).

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. We carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of ≈ 3 × 10 − 7 dpa/s at temperatures between 300 and 800 K. [ABSTRACT FROM AUTHOR]

Iron-Chromium-Aluminum (FeCrAl) alloys are candidate materials for the cladding of light water reactor (LWR) fuels. The FeCrAl alloys in general range in Cr composition from 12% (C26M) to 21% (APMT). In this work, the general corrosion behavior of Additively Manufactured (AM) C26M coupons was compared to the behavior of traditional Powder Metallurgy (PM) coupons. Immersion testing were conducted for 12 months at 288 °C and 330 °C in pure water containing either oxygen or hydrogen. Results show that the mass change of AM specimens in hydrogenated water was like the mass change of PM specimens. In oxygenated water, the mass change of AM coupons was higher and less reproducible than for the PM coupons. Porosity in the AM specimens makes their behavior less predictable in high-temperature water. [ABSTRACT FROM AUTHOR]

Accident tolerant fuel (ATF) cladding is a new type of nuclear fuel cladding designed to improve the safety and performance of nuclear reactors. In this paper, the kinetics and degradation mechanisms during high-temperature oxidation in steam of the three most promising ATF cladding materials, i.e., chromium-coated zirconium alloys, FeCrAl alloys, and silicon carbide-based composites, are described. Each system has its own degradation mechanisms leading to different maximum survival temperatures. After providing general information and data to understand the oxidation and degradation processes, illustrative examples obtained at the Karlsruhe Institute of Technology are given for each type of cladding. The maximum temperatures at which the barrier effect of the cladding can be maintained for a reasonable period of time during nuclear accident scenarios are 1200–1300 °C for Cr-coated Zr alloys, 1400 °C for FeCrAl alloys, and 1700 °C for SiC-based composite claddings. [ABSTRACT FROM AUTHOR]

This work aims to analyze the passivation and pitting corrosion behaviors of laser beam welded R60702 zirconium alloy in neutral and acidic chloride‐containing electrolytes. Potentiodynamic polarization and electrochemical impedance spectroscopy measurements are carried out to investigate the electrochemical performance of the welding joint. Scanning electron microscope, X‐ray diffraction, and three‐dimensional profile digital microscope are utilized to reveal the microstructures and corrosion morphologies. The electrochemical results show that the heat‐affected zone (HAZ) and weld zone have nearly equal corrosion performance, and both of them are more corrosion‐resistant than the base metal. The corrosion morphologies suggest that the HAZ has the lowest sensitivity to pitting corrosion. Moreover, it is unraveled that, in chloride‐containing electrolytes, the quantity and distribution of Zr(Fe,Cr)2 particle phases are the main factors that determine the different corrosion performances of the welding joint. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Zirconium alloy with the composition of 90.1% Zr and alloying elements such as Mg, Al, and Si was investigated for its mechanical properties and corrosion resistance. The specimen was dissolved in a mixture of NaOH, Na2CO3, S, and SnO2 at 800°C for 10 min, followed by an HCl solution at room temperature for 100 cycles. The structural properties were characterized by X-ray diffraction and scanning electron microscopy, while the mechanical properties, such as tensile strength were investigated by a universal tensile machine. The microstructural observations indicated that the outside part of the crucible underwent pitting corrosion which resulted in corrosion in all directions inside that part of the crucible. The corrosion structure consisted of cracks that reached the base metal. The residual chemical from the fusion and dissolution process remained in the crack and formed NaCl, which accelerated the crack. The release rate of the zirconia oxide layer was calculated to be 8 μm per cycle. During the corrosion process at high temperatures, oxygen diffusion infiltrated the base metal and stretched the crystal lattice, causing strain hardening with the values of yield strength and ultimate tensile strength increased to 22.5% while the strain decreased by 50%. [ABSTRACT FROM AUTHOR]

In this study, triple sheets of R60702 zirconium (Zr) alloy are joined together using the resistance‐spot‐welding process. The nugget formation mechanism and microstructure of the weld joint in three equal‐thickness sheets of R60702 Zr alloy are studied. In the results, it is revealed that the initial nugget forms at the top of the intermediate sheet, and subsequently expands both radially and axially until a complete nugget is formed. The fusion zone (FZ) exhibits a martensitic microstructure, while the heat‐affected zone (HAZ) consists of two regions: coarse‐grain HAZ and fine‐grain HAZ. The hardness in the FZ is higher than that of the HAZ and the base metal. Furthermore, the tensile‐shear force and fracture mode of the weld nuggets are also studied. The weld joints exhibit a maximum tensile‐shear load of 5.7 kN under the following welding parameters: welding current of 14 kA, welding time of 150 ms, and welding force of 3.6 kN. Three failure modes are identified interface fracture, partial button fracture, and button fracture. [ABSTRACT FROM AUTHOR]