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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]

Research on accident-tolerant fuels (ATFs) for Light Water Reactors (LWRs) has focused on improving the safety of zirconium alloy fuel rods since the Fukushima accident in 2011. Additionally, normal LWR operating conditions cause fretting wear on the fuel rod surface, which requires tribological improvements. Chromium-based coatings on zirconium alloys are one of the most advanced concepts for developing a fuel that can withstand accidents. Both under operating conditions and when subjected to high-temperature steam, Zr alloys with protective coatings demonstrated improvements, compared to uncoated surfaces. This study aims to evaluate the effect of substrate bias on the structure and tribological properties of chromium films deposited by magnetron sputtering onto Zr alloy nuclear fuel rod cladding. The coatings were also characterised by scanning electron microscopy (SEM), energy dispersive X-ray spectroscopy (EDX), X-ray diffraction (XRD), atomic force microscopy (AFM), transmission electron microscopy (TEM) and optical profilometry. The hardness of the coating was measured with a nano-indenter. Initial finding indicates that the application of a low substrate bias of −50 V enhances film density and adhesion and also reduces the coating wear rates, in comparison to higher biases (−100 V to −150 V) or no applied bias. • Chromium coatings on zirconium alloy were prepared by magnetron sputtering with bias control. • A −50 V substrate bias refines the coating structure, increasing density and wear resistance. • Optimized bias enhances chromium coating adhesion and wear resistance for protective properties. [ABSTRACT FROM AUTHOR]

• With the help of well-trained moment tensor potential, the phonon-level physical mechanism of how Nb and Sn doping alters the lattice thermal conductivity of zirconium is elaborated, which has positive significance for developing advanced accident-tolerant fuel candidate Zircaloys. • Nb and Sn doping leads to a significant reduction in the value and anisotropy of lattice thermal conductivity of zirconium. • The reduced average phonon group velocity and enhanced phonon-phonon scattering are the primary factors of the decrease in lattice thermal conductivity of zirconium alloys after Nb and Sn doping. • The addition of Nb and Sn alloying elements increases the anharmonic phonon-phonon scattering rate and high-frequency phase space of phononic emission process, which contributes to larger three-phonon scattering probability. Zirconium alloy coating is an important direction for the modification of nuclear cladding materials. Thermal conductivity is a critical property of cladding materials. With extensively studying phonon-electron non-equilibrium energy transfer processes in the thermal transport of zirconium alloy coating, to distinguish the contributions from phonon and electron thermal conductivity of Zr alloys becomes crucial and necessary. In this work, we successfully predicted the lattice thermal conductivities of zirconium, Zr-Sn and Zr-Nb using machine learning potentials. Sn and Nb doping leads to a significant decrease in lattice thermal conductivity, which is mainly due to the alterations in phonon group velocity and phonon scattering. The larger atomic mass of doping elements and weakened interatomic interactions of Zr-Nb together lead to a significant decrease in phonon group velocity. Doping Sn and Nb also increases phonon-phonon scattering rate and three-phonon scattering channels, resulting in a shortening in phonon lifetime and a decrease in lattice thermal conductivity. [ABSTRACT FROM AUTHOR]

• Irradiation-induced defects were observed in the Zr-1NB alloy subjected to neutron irradiation in a VVER-1000 type reactor for 4 years of operation. • Increase in the yield strength is attributed to the presence of < a >-type loops in the material. • Deformation occurred by both dislocation channeling (slip) and twinning. • { 01 1 ¯ 1 } 〈 0 1 ¯ 12 〉 type compression twinning was observed after deformation at 350 °C. In this study, a detailed investigation was carried out on the recrystallized Zirconium (Zr-1Nb) alloy, irradiated in the reactor core of Temelín NPP (VVER-1000) for a duration of four annual cycles within the framework of the ALVEL-ČEZ project. The main objective of the present work is to report the microstructural evolution, with a particular focus on the radiation-induced defects and elucidate the underlying deformation mechanisms. The combined use of electron backscattered diffraction and scanning transmission electron microscopy enables to obtain the statistical details of the microstructural changes in the irradiated material. To assess the deformation behavior of the irradiated Zr-1Nb alloy, tensile testing was conducted at room temperature and 350 °C and a strain rate of 2 × 10-3 s-1 along the axial direction of the tube. Transmission electron microscopy observations on the deformed sample revealed the presence of dislocation-free channels within the grains. Based on the stereographic trace analysis, it was determined that slip occurred through prismatic, pyramidal and basal channels. Additionally, the { 01 1 ¯ 1 } 〈 0 1 ¯ 12 〉 type twinning system was also found to be activated in the deformed sample. [ABSTRACT FROM AUTHOR]

The formation of the Zr-Cr eutectic (∼1320 °C), which does not occur in conventional Zirconium alloys, introduces a significant safety concern for Cr-coated Accident Tolerant Fuel (ATF). This study investigated the Peak Cladding Temperature (PCT) limit for Cr-coated ATF by examining the effects of the Zr-Cr eutectic on the mechanical integrity of Cr-coated Zr-1.1Nb claddings. To achieve this, Integral Loss of Coolant Accident (LOCA) tests and Ring Compression Tests (RCTs) were conducted on Cr-coated specimens under both steam and oxygen-free environments. The results indicated that while the formation of the eutectic phase between Zr and Cr does not result in structural failure, it reduced the ductility of the cladding. However, the impact of Zr-Cr eutectic on the reduction in ductility was overshadowed by the significant impact of the oxidation under the same conditions. The primary cause of the severe ductility loss in specimens oxidized above the eutectic onset temperature was the increased oxygen diffusion at elevated temperatures. Consequently, compared to specimens oxidized at 1204 °C, the increased oxygen concentration in the ductile layer further reduced the ductility of the cladding. Based on these findings, the pronounced reduction in ductility caused by oxidation of the Zr matrix in Cr-coated ATF cladding underscored the necessity of adhering to the current PCT limit, as long as the cladding matrix of Cr-coated ATF cladding remains Zirconium alloy. Furthermore, the excessive embrittlement observed in Zirconium alloy at temperatures above 2400 °F (1315 °C) was a key factor in establishing the current 2200 °F (1204 °C) PCT limit. As a result, extending the PCT limit beyond 1204 °C for Cr-coated ATF cladding is impractical, given the rapid oxygen diffusion and the consequent reduction in ductility at these higher temperatures. Therefore, maintaining the current 2200 °F (1204 °C) PCT limit for Cr-coated ATF cladding can serve as the effective approach for ensuring the safety of Cr-coated ATF cladding within the existing regulatory framework. [ABSTRACT FROM AUTHOR]

For nuclear fuel elements, the interface mechanical properties of zirconium alloy cladding is critical to the safety and reliability of reactors. However, due to the small thickness of the fuel plates (<2 mm), accurately capturing the behavior of interface cracks is challenging, complicates the measurement of interface fracture toughness. This study developed a data-driven inversion method using the generalized regression neural network (GRNN) to rapidly and accurately determine the fracture toughness of zirconium alloy cladding interface. The database was established by combining splitting experiments with numerical simulations. The cohesive zone model was utilized to accurately simulate crack propagation paths and fracture modes in numerical simulations. The influences of key parameters such as cohesive strength, stiffness, and interface fracture energy were analyzed in detail. After extensive training, the prediction model accurately forecasted the interface fracture toughness. The results indicate that the proposed GRNN-based inversion approach is feasible and effective for predicting the fracture toughness of zirconium alloy cladding interface, and can be extended to determinations of other mechanical properties in the nuclear fuel element. [ABSTRACT FROM AUTHOR]

Plastic anisotropy is observed during plastic flow in samples from the axial and transverse directions of Zr-2.5Nb pressure tubes used in CANDU 1 1 CAN ada D euterium U ranium - Pressurized Heavy Water power reactors. nuclear reactors tested in uniaxial tension. Plastic anisotropy was also measured in shear by testing in torsion 'mini' pressure tubes from the same material. Room temperature results from these experiments were analysed using a visco-plastic self-consistent model that takes into account the crystallographic texture of the material and allows for the single crystal work hardening behaviour to be described by means of a deformation law specific to each slip system. The visco-plastic self-consistent model was used to derive: (i) the evolution with strain of the critical resolved shear stress values consistent with prismatic, basal and pyramidal dislocation glide, (ii) the evolution of slip system activity as a function of strain, and (iii) the values of Hill's plastic anisotropy coefficients that are consistent with the observed anisotropy of yielding and their dependence on accumulated strain. The model also allows for the prediction of the components of the flow stress tensor that cannot be measured experimentally and their dependence on the work hardening behaviour of pressure tube material. Moreover, the yield surface after different amounts of plastic strain was calculated using the visco-plastic self-consistent model, compared to the one that results from using the Hill's anisotropy coefficients. Our work exposes the limitations of the Hill ellipsoid for describing plastic yield when microstructural evolution is present. [Display omitted] [ABSTRACT FROM AUTHOR]

The study examined the diffusion mitigation effect and mechanical properties of the Nb interlayer prepared by magnetron sputtering as a Cr Zr diffusion barrier through annealing experiment, high-temperature air oxidation experiment and ring compression tests. The results show that the presence of an Nb interlayer significantly alleviates the Cr Zr interdiffusion, especially prevents the Cr Zr interdiffusion in the 1200 °C air environment and stabilizes the interface structure in the 1400 °C air environment. After undergoing air oxidation at 1200 °C, the Cr/Nb-coated samples showed the structure of Cr 2 O 3 , Cr, Cr 2 Nb Laves phase, Nb, (Zr, Nb) miscibility gap, β-Zr, α-Zr(O), and ZrO 2 layers. During a certain period of argon annealing at 1332 °C, the increase in the thickness of the interlayer of the Cr/Nb coating, especially for samples containing micron sized Nb layer, is smaller than that of Cr-coated samples, indicating that the presence of the Nb layer is effective in alleviating the Cr Zr contact reaction to a certain extent. During the ring compression tests, the Cr/Nb coating is still firmly bonded to the Zr alloy cladding tube even under large deformation, and the micron-scale Nb layer releases part of the stress, resulting in short crack clusters on the surface of the coating instead of a large number of through-wall cracks. In summary, the Nb layer has the potential to be used as a diffusion barrier for Cr Zr interface, and the details of Nb as a diffusion barrier are discussed. • The Nb interlayer was prepared by magnetron sputtering as a Cr Zr diffusion barrier layer. • The Nb interlayer restrain the interdiffusion of Cr Zr. • The addition of Nb layer increases the melting point of the coating and improves the stability of the interfacial structure. • The Cr/Nb coating and the zirconium alloy substrate are still well bonded under stress. [ABSTRACT FROM AUTHOR]

• Cr coatings with different columnar structure were constructed by Bipolar-HiPIMS. • Cr coatings with dense & column-oblate structure possess high hardness value. • Optimal Cr coatings with dense structure can withstand 2 h steam oxidation at 1200 °C. • 1200 °C steam oxidation mechanism of ameliorated Cr coatings is discussed. In order to improve the 1200°C steam oxidation properties of Cr-coated zirconium alloy accident tolerant fuel cladding tubes, chromium coatings with various micro-structures were fabricated by bipolar high-power impulse magnetron sputtering technology. The microstructures of chromium coatings transforms from unconsolidated column structure to dense & column-oblate structure with the bias voltage increase. Hardness and elastic modulus values of optimal chromium coatings are approximately 19.1 GPa and 374.8 GPa, respectively, which are 2.09 and 1.60 times higher than those of chromium coatings with unconsolidated column structures. Optimal chromium coatings with dense & column-oblate structures display preferable 1200°C steam oxidation resistance because of possession lower weight gain value and thicker Cr 2 O 3 oxide layer after 2 h exposure periods. The fabrication strategy of chromium coatings with dense & column-oblate structure is expected to pave the way for enhancing 1200 °C steam oxidation resistance for Cr-coated zirconium alloys. [ABSTRACT FROM AUTHOR]

Cr 3 Si and Cr coatings were deposited on Zry-4 substrates via magnetron sputtering to examine their oxidation behavior in steam ranging from 1200 to 1400 °C. At 1200 °C, a dense Cr 2 O 3 layer that formed on the surface of the Cr 3 Si-coated Zry-4 followed a power-law function, preventing the internal diffusion of O. Additionally, an in-situ Zr 2 Si layer, developed between the Cr 3 Si coating and Zry-4 substrate, significantly mitigated the outward diffusion of Zr. However, at 1330 °C in steam, the continuity of the Zr 2 Si layer was disrupted due to the formation of ZrCr 2 , which facilitated the outward diffusion of Zr to the interface between the remaining Cr 3 Si layer and the outer Cr 2 O 3 layer. Despite the reduction in the Cr 2 O 3 layer thickness caused by redox reactions between Cr 2 O 3 and Zr, the Cr 3 Si coating still managed to delay the steam oxidation of the Zry-4 substrate for over 60 min at 1330 °C. While on the uncoated side, the thickness of the formed ZrO 2 reached 846 ± 5.2 μm. For the Cr coating, after 30 min of oxidation in steam at 1330 °C, a significant liquid eutectic formed between the Cr coating and the Zry-4 substrate, compromising the coating's overall structural stability. The oxidation resistance of the Cr 3 Si coating surpasses that of the Cr coating after exposure to steam at 1400 °C for 10 min. These findings suggest that the Cr 3 Si coatings hold promise for robustly protecting zirconium alloy cladding in scenarios involving design basis accidents (DBA) and beyond design basis accidents (BDBA). • A continuous Zr 2 Si can effectively inhibit the interdiffuse of Cr and Zr, improving the stability of the Cr 3 Si-coated Zry-4 system. • In steam above 1330 °C, the intense diffusion of Cr and Zr disrupts the continuity of the Zr 2 Si layer, reducing the oxidation resistance of the Cr 3 Si coating. • In 1200 °C steam, the growth kinetics of Cr 2 O 3 on the surface of the Cr 3 Si coating follows a power-law function. [ABSTRACT FROM AUTHOR]

Hydrogen-induced variations in mechanical behavior of zirconium alloys impose detrimental influence on nuclear fuel cladding integrity. This work reports a disappearance of intrinsic yield drop in a recrystallized zirconium alloy following hydrogen-charging treatment. Microstructure characterizations reveal that the nano-hydrides precipitation, mediated by second phase particles Zr(Fe,Cr) 2 acting as hydrogen trapping sites, leads to emission of substantial dislocations in α-matrix grains due to strong strain concentrations, as identified by high-angular resolution EBSD. These mobile dislocations preserved at elevated temperatures can maintain the applied plastic strain and impede rapid dislocation multiplication as well, a conclusion validated by comparative analysis of dislocation densities prior to and near yielding stage. These findings are expected to shed light on the underlying mechanisms governing the interaction between hydrogen and microstructural defects in Zr-based nuclear fuel cladding materials. [Display omitted] [ABSTRACT FROM AUTHOR]

This study compared the uniform corrosion behavior of Zr-Sn-Fe-Cr-Ni alloys in deionized water at 360 °C/18.6 MPa with high and low concentration of dissolved oxygen (DO). It was found that high concentration of DO accelerated the corrosion rate of the Zr-Sn-Fe-Cr-Ni alloys and led to an earlier corrosion transition. Increased DO concentration resulted in a higher content of t-ZrO 2 near the metal-oxide interface, which induced greater in-plane compressive stress in the oxide film and a high-level phase transformation from t-ZrO 2 to m-ZrO 2. This, in turn, led to an earlier occurrence of corrosion transition. [ABSTRACT FROM AUTHOR]

This paper describes a series of experiments conducted to investigate the in-situ irradiation creep behavior of Zircaloy-4 using proton irradiations as a surrogate to neutrons. The first series of experiments investigated the impact of the initial irradiation-induced defect evolution during the 0 – 0.1 dpa regime on the subsequent in-situ steady-state behavior derived from experimentation. These experiments were conducted at a constant applied load of 85 MPa, a constant damage rate of 1.58 × 10−6 dpa/s, and were repeated at both 250 °C and 350 °C. We also examined the use of a 'conditioning-irradiation' step prior to the creep tests on the results derived from subsequent in-situ proton irradiation creep experiments. By extension, we aimed to further develop and refine optimum testing procedures when using proton irradiations to investigate the in-situ creep behavior of nuclear materials. The second series of in-situ proton irradiation experiments were conducted on two Recrystallized (RX) Zircaloy-4 samples in order to investigate the temperature and stress dependence of irradiation creep. One sample was held at a constant load while the temperature was varied in the range of 275 – 350 °C, and the other was held at a constant temperature while stress was varied in the range of 65 – 105 MPa. The associated strains and creep rates were measured at each interval and used to determine an activation temperature of 5000 ± 1700 K and a stress sensitivity exponent of 4.3 ± 0.8 for RX Zr-4 over the given temperature and stress ranges. A discussion of potential deformation mechanisms based on competition between bulk diffusion through the lattice and point defect diffusion enabling dislocation climb and glide is given: the relatively high stress dependence suggested the latter is more likely however further investigations will be required to improve the mechanistic understanding. The results presented in this manuscript, including activation temperature, stress sensitivity, and associated creep rates, determined through proton irradiation investigations are closely comparable to those determined from neutron irradiation experiments found in the literature for similar zirconium alloys at similar temperatures and stresses. Coupled with the primary-secondary creep behavior investigations presented, this demonstrates the usefulness of this approach to estimate neutron-irradiation equivalent creep behavior using in-situ proton irradiation. [ABSTRACT FROM AUTHOR]

The tribocorrosion behavior of the three representative dissolver materials, namely zirconium (Zr) alloy, 310 stainless steel (SS), and titanium (Ti) alloy in HNO 3 was systematically investigated by SEM, XPS, and Nanoindentation. The results revealed that Zr alloy possesses high corrosion resistance but a less favorable tribocorrosion resistance, whereas 310 SS presents a comparatively better tribocorrosion resistance but insufficient corrosion resistance. Moreover, both the coefficient of friction and wear loss of all three alloys were decreased at 0.8 V compared to −0.8 V, demonstrating an antagonistic effect of corrosion on wear, attributed to the lubrication and hardness enhancement of oxide layers. The present work offers a new perspective on the tribocorrosion behavior of continuous dissolvers used for reprocessing spent nuclear fuel. [Display omitted] • Tribocorrosion behavior of Zr alloy, 310 SS, and Ti alloy in HNO 3 was studied. • Zr alloy is high corrosion resistant but has insufficient tribocorrosion resistance. • 310 SS presents good tribocorrosion resistance but poor static corrosion resistance. • Antagonistic effect between corrosion and wear occurs on three alloys. • A tribocorrosion model of three dissolver materials in HNO 3 was proposed. [ABSTRACT FROM AUTHOR]