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Showing papers on "Zirconium alloy published in 2022"


Journal ArticleDOI
TL;DR: In this paper , the oxidation, diffusion, and mechanical properties of Cr-coated Zr alloys in normal operation conditions and accident conditions of nuclear reactors are reviewed, and the factors that cause the failure of the coating are analyzed, and some questions that need to be clarified and further studied are proposed.

58 citations


Journal ArticleDOI
TL;DR: In this article, the oxidation, diffusion, and mechanical properties of Cr-coated Zr alloys in normal operation conditions and accident conditions of nuclear reactors are reviewed, and the factors that cause the failure of the coating are analyzed, and some questions that need to be clarified and further studied are proposed.

58 citations


Journal ArticleDOI
TL;DR: In this article , the performance of two types of chromium-coated zirconium alloys (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on prototype rod segment samples filled with ZrO2 pellets and tightly sealed with welded end caps.

23 citations


Journal ArticleDOI
TL;DR: In this article , a single-layer Cr (8 um) and multilayer Cr(8 μm)/Mo (3 μm) coatings were deposited by magnetron sputtering.
Abstract: The oxidation and interdiffusion behavior of Cr- and Cr/Mo-coated Zr1Nb zirconium alloy were investigated. A single-layer Cr (8 um) and multilayer Cr (8 μm)/Mo (3 μm) coatings were deposited by magnetron sputtering. The coated Zr1Nb alloy samples were oxidized in air at 1100 °C for 15–60 min. Both coating types had a protective scale during high-temperature (HT) oxidation. The use of Mo sublayer resulted in preventing CrZr interdiffusion under high temperature. Scanning electron microscopy, optical microscopy, X-ray diffraction before and after HT oxidation and in situ X-ray diffraction up to 1250 °C were used to identify interdiffusion behavior of the system of Cr/Mo/Zr. Some aspects of applying of Cr/Mo coatings for Zr nuclear fuel claddings are discussed.

15 citations



Journal ArticleDOI
TL;DR: In this paper , the microstructure and oxidation behavior of Cr coatings on the zircaloy-4 in steam at 1300 °C were studied systematically, and they showed improved steam oxidation over ziraloy up to 1300°C.

14 citations


Journal ArticleDOI
TL;DR: In this article , the authors compared the differences in the hydrogen-induced cladding embrittlement of cold work stress-relief annealed (CWSR or SRA) Zircaloy-4 and Zr-Nb alloy cladding with ring compression test at the temperature of the spent fuel pool, which is approximately 40 °C.

14 citations


Journal ArticleDOI
TL;DR: In this paper , Mo and Al2O3 powders were successfully coated on Zircaloy-4 by air plasma spraying and their morphologies, phases of corrosion products and hardness were analyzed.

14 citations


Journal ArticleDOI
TL;DR: In this article, a post-Loss of Coolant Accident (LOCA) ductility assessment of Cr-coated Zircaloy cladding was conducted in compliance with the United States Nuclear Regulatory Commission (U.S.NRC)’s guidelines.

11 citations


Journal ArticleDOI
TL;DR: In this paper, a 15 µm-thick chromium coating on a zirconium alloy substrate using a particular physical vapor deposition process is studied at room temperature using several experimental techniques at different scales: biaxial tests (internal pressure + axial tension) with several stress biaaxiality ratios on outer-coated tubes; in situ tensile tests in a scanning electron microscope (SEM) on coated sheet samples.

10 citations


Journal ArticleDOI
TL;DR: In this paper , the rapid deposition of a thick chromium coating on zircaloy-4 cladding tubes by atmospheric plasma spraying was studied, and the nanohardness and modulus of the sprayed chromium coatings were higher than those of the substrate.

Journal ArticleDOI
TL;DR: In this article, a series of high-temperature steam oxidation conditions were used to study the damage evolution of Zr alloy claddings under different conditions, and the results showed that a large proportion of cracks would self-healed when the temperature increased to 1000 ˚C.

Journal ArticleDOI
TL;DR: In this article , cracks were initially produced within the Cr coat by internal pressure creep test and the Cr-coated samples with cracks were then oxidized at temperatures from 800 °C to 1200 °C in water steam environment.

Journal ArticleDOI
TL;DR: In this paper , a novel mechanism of orientation-dependent Cr-coating corrosion was demonstrated, which is expected to guide optimized design of cr-coated nuclear fuel cladding material.

Journal ArticleDOI
TL;DR: In this article , a review of the current Zr alloys and opportunities for advanced zirconium alloys to meet the demands of a structural material in fusion reactors is presented.

Journal ArticleDOI
TL;DR: In this paper , the evolution of irradiation defects with irradiation is taken into account, especially to deduce the local growth strain, and a good description of the in-reactor behavior is obtained with the irradiation defect evolution consistent with Transmission Electron Microscopy observations.

Journal ArticleDOI
TL;DR: In this article , the authors analyzed fractions of each zirconium-hydride interface orientation relationship by EBSD (Electron Backscatter Diffraction) and found that the typical macroscopic radial hydrides are primarily formed upon aggregation of {10 1 ¯ 1} α-Zr // {111} δ − ZrH1.66 and {0001} α −Zr − δ -ZrH 1.66 orientation relationships in mesoscale.

Journal ArticleDOI
TL;DR: In this paper , a simple planar dislocation model was developed of the hydride-dislocation ensemble in α-Zr and a discrete dislocation plasticity was used to model the diffuse plastic relaxation associated with hydrides formation.
Abstract: The precipitation of hydrides in zirconium alloys is accompanied by a significant and anisotropic volumetric expansion. Previous literature quantified the misfit both theoretically and experimentally, but these values differ greatly; the experimental values are consistently lower. One possibility is that the experimental measurements include the effect of dislocations generated by the hydride, which relax the transformation stresses. To test this hypothesis, it is important to determine the stress field of a hydride and its associated dislocations, combined. A simple planar dislocation model was developed of the hydride—dislocation ensemble in α-Zr. By capturing details of the dislocation structures given in the literature, it is shown in this study that including the interfacial dislocations largely reconciles the predicted and experimental values. Discrete dislocation plasticity is then used to model the diffuse plastic relaxation associated with hydride formation. The effects of plastic relaxation on the equilibrium hydrogen profile, hence the implications for subsequent hydride precipitation, are discussed. In particular, precipitation–dissolution cycles were simulated to calculate the magnitude of the residual hydrostatic tension, which is argued to be the primary cause of the “memory effect” for the re-precipitation of both γ and δ hydrides.

Journal ArticleDOI
TL;DR: In this paper , the effects of hydride amount (20 −850 wppm), orientation (circumferential and radial), and temperature (room temperature, 100 °C, 200 °C) on the axial mechanical properties of Zircaloy-4 cladding were comprehensively examined.

Journal ArticleDOI
TL;DR: In this paper , the corrosion behavior of FeCrAl alloy and Zircaloy-4 coated with NiCr in hydrogenated water was studied. And the results showed that NiCr coating was well-covered on the Ziraloy 4 substrate after exposure.

Journal ArticleDOI
TL;DR: In this article, APT and TEM analyses were used to characterise two Zr-based alloys, Zircaloy-2 and low-Sn ZIRLO, for predicting changes in the properties of Zr alloys in-service in fission reactors.

Journal ArticleDOI
TL;DR: In this paper , the authors used a stored energy density fracture criterion to simulate crack propagation in single crystals and polycrystalline microstructures for comparison with experimental crack growth rate data.

Journal ArticleDOI
TL;DR: In this article , the effect of boron and zirconium on the microcracking susceptibility of the Ni-base superalloy IN-738LC during laser powder bed fusion (LPBF) was studied using custom designed powder grades.

Journal ArticleDOI
TL;DR: In this paper, the authors presented the limitation of the HNGD model and proposed two hypotheses to improve the model's accuracy, which can be used in Bison to model Zircaloy cladding with a zirconium inner liner.

Journal ArticleDOI
TL;DR: In this article , APT and TEM analyses were used to characterise two Zr-based alloys, Zircaloy-2 and low-Sn ZIRLO, for predicting changes in the properties of Zr alloys in-service in fission reactors.

Journal ArticleDOI
TL;DR: In this paper , burst and steam oxidation behavior of bare and Cr-coated Zirlo claddings were examined under simulated loss-of-coolant accident conditions, and the 4.4 µm coating had no substantial effect on ballooning or opening geometry but did increase burst temperatures at higher pressures.

Journal ArticleDOI
TL;DR: In this article , the authors presented the limitation of the HNGD model and proposed two hypotheses to improve the model's accuracy, which can be used in Bison to model Zircaloy cladding with a zirconium inner liner.

Journal ArticleDOI
TL;DR: In this article , the authors studied the hydrogen diffusion and precipitation patterns during delayed hydride cracking (DHC) as a function of temperature via high-resolution neutron imaging and found that the elevated local concentrations correlate with the amount of hydrogen in solid solution available to diffuse towards the crack tip as well as the tensile stresses at the crack.

Journal ArticleDOI
28 Oct 2022-Energies
TL;DR: In this paper , the authors evaluated the neutronics performance of a chromium coating concept and design solutions for a Zircaloy-uranium fuel system (Zr-U) and a two-dimensional full core based on an APR-1400 reactor design.
Abstract: The accident-tolerant fuel concept involves replacing the conventional cladding system (zirconium) with a new material or coating that has specific thermomechanical properties. The aim of this study is to evaluate the neutronics performance of a chromium coating concept and design solutions. A Zircaloy–uranium fuel system (Zr–U) is currently used as a standard fuel system in pressurized water reactors around the world. This investigation presents the benefits of utilizing an alternative cladding material such as chromium coating and the effects on the thermal neutron parameters of the way in which the chromium coating is introduced in the reactor fuel. Among these significant benefits is an increase in the reactor fuel’s thermal conductivity, which improves reactor safety. Two types of fuel-cladding systems were investigated: Zircaloy–uranium (Zr–U) and Zircaloy–chromium (Zr–Cr–U) coating fuel systems. Neutronics analysis evaluations were performed for the selected fuel assemblies and a two-dimensional full core based on an APR-1400 reactor design. Neutronics analyses were performed for the application of the new fuel-cladding material systems using the reactor-physics Monte Carlo code Serpent 2.31.

Journal ArticleDOI
TL;DR: In this article , the effect of zirconium content on the mechanical properties of ODS FeCrAl alloys, a kind of promising cladding materials for advance nuclear energy system, was investigated.
Abstract: The aim of this work is to investigate the effect of zirconium content on the mechanical properties of ODS FeCrAl alloys, a kind of promising cladding materials for advance nuclear energy system. Four 13Cr5Al-ODS alloys with different Zr contents (0, 0.3, 0.6 and 1.2 wt%, respectively) were fabricated by mechanical alloying (MA) and hot isostatic pressing (HIP). The mechanical properties at different temperatures were measured, then microstructures were characterized and compared by SEM and TEM to figure out the relationship between composition and mechanical properties. The results show that when adding 0.6 wt% Zr, the sample shows the highest number density of dispersed particles, reaching 1.103 × 1023/m3. Furthermore, its room temperature tensile strength reaches 1012.94 MPa while maintaining a high elongation of 16%. Likewise, the strength advantage of this sample is more obvious at high temperature.