The welding joints of Zircaloy 4 plates obtained by diffusion welding at 800°C under pressure in vacuum were cracked during autoclave tests at 400°C superheated steam after exposure longer than 150 days. T...The welding joints of Zircaloy 4 plates obtained by diffusion welding at 800°C under pressure in vacuum were cracked during autoclave tests at 400°C superheated steam after exposure longer than 150 days. The section of specimens was examined by optical microscopy and the composition at the tips of cracking was analyzed by electron microprobe. The result shows that the combination of oxidation and hydriding induced cracking is responsible for this failure of the welding joints.展开更多
In order to investigate the effect of lanthanum ion implantation on theoxidation behavior of zircaloy at 500℃, Zircaloy specimens were implanted by lanthanum ions with adose range from 5xl0^(16) to 2xl0^(17) ions/cm^...In order to investigate the effect of lanthanum ion implantation on theoxidation behavior of zircaloy at 500℃, Zircaloy specimens were implanted by lanthanum ions with adose range from 5xl0^(16) to 2xl0^(17) ions/cm^2 at room temperature, and then oxidized at 500℃ for100 min. The valence of the oxides in the scale was analyzed by X-ray Photoelectron Spectroscopy(XPS). The phase structures of the oxides in the scale were examined by Glancing Angle X-rayDiffraction (GAXRD). With the increase of implanted lanthanum ions dose, the phase structures in theoxide scale are transformed from monoclinic zirconia to hexagonal one and then to monoclinic oneagain. The measurement of weight gain showed that a similar change from the decreased gain toincreased one again is achieved in the oxidation behavior of lanthanum ion implanted zircaloycompared with that of as-received zircaloy.展开更多
In this work, hydrogen absorption and the permeation behavior of the passive layer formed on zircaloy-4 are in- vestigated. Potentiodynamic polarization, Mott-Schottky analysis, electrochemical impedance spectroscopy,...In this work, hydrogen absorption and the permeation behavior of the passive layer formed on zircaloy-4 are in- vestigated. Potentiodynamic polarization, Mott-Schottky analysis, electrochemical impedance spectroscopy, and Raman scattering spectroscopy are employed to characterize the passive defects before and after hydrogen permeation. It is found that the nanoscale passive ZrO2 films play an important role in the resistance against corrosion; hydrogen impingement, however, reduces the passive impedance towards hydrothermal oxidation. The increase of defects （vacancies） in passive film is probably attributed to the degradation. We believe that this finding will provide valuable insight into the understanding of the corrosion mechanism of zircaloys used in light water reactors.展开更多
Fatigue dislocation configurations of Zircaloy-4 at 470℃×1h stress-relieved condition and 620℃×1h recrystallized condition were analyzed using TEM. Theresults show that: {1 0 1 0} prismatic slip is the pri...Fatigue dislocation configurations of Zircaloy-4 at 470℃×1h stress-relieved condition and 620℃×1h recrystallized condition were analyzed using TEM. Theresults show that: {1 0 1 0} prismatic slip is the primary deformation mode at RT. Prismatic and pyramidal slips are activated simultaneously at 400℃. The typicalsubstructure is the elongated dislocation lines at RT; whereas at 400℃, it is rectangularcells in stress-relieved specimens, and elongated cells plus dipole perpendicular cellboundary in recrystallized specimens. The relationship map among dislocation configuration, test temperature and cyclic strain range is established, finally.展开更多
The contrastive corrosion experiments between surface nanocrystallined Zircaloy-4 and coarse-grained Zircaloy-4 under the condition of 673 K/10.3 MPa in pure water are carried out, and the microstructure of oxide film...The contrastive corrosion experiments between surface nanocrystallined Zircaloy-4 and coarse-grained Zircaloy-4 under the condition of 673 K/10.3 MPa in pure water are carried out, and the microstructure of oxide films has been studied. The results indicate that the growth rate of oxide films formed on the nanocrystalline Zircaloy-4 is lower than that of oxide films formed on the coarse-grained Zircaloy-4. Simultaneously, the oxide/metal interface of the former is more regular and glossy than that of the latter. For nanocrystalline Zircaloy-4, the low oxygen diffusion rate through the oxide/metal interface can hinder the reaction of oxygen ion with metal ion. Furthermore, more tetragonal ZrO2 are observed in the oxide films, which can delay the martensite phase transition from tetragonal to monoclinic phase in oxide films.展开更多
After being treated in different ways, Zr-Sn-Nb-Fe alloy specimens are exposed in 0.01mol/L LiOH aqueous solution at 350℃ under 16.8 MPa. The examination of microstructures and second phase particles （SPPs） of thes...After being treated in different ways, Zr-Sn-Nb-Fe alloy specimens are exposed in 0.01mol/L LiOH aqueous solution at 350℃ under 16.8 MPa. The examination of microstructures and second phase particles （SPPs） of these specimens was carried out by high-resolution transmission electron microscopy （HR-TEM）. The specimens treated at 800℃ before the final cold rolling have a better corrosion resistance than those treated at 680℃, and the specimens treated at 500℃, after the final cold rolling, have a better corrosion resistance than those treated at 560℃. TEM examination shows that the SPPs existing in the 800℃/500℃ specimen, which has the best corrosion resistance, contains a lot of Nb element, which results in the reduction of the niobium content in the α-Zr solid solution.展开更多
In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as...In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as cladding material mainly due to its lower neutron absorption cross section. Now, stainless steel cladding appears as a possible solution for safety problems related to hydrogen production and explosion as occurred in Fukushima Daiichi accident. The aim of this paper is to discuss the steady-state irradiation performance using stainless steel as cladding. The results show that stainless steel rods display higher fuel temperatures and wider pellet-cladding gaps than Zircaloy rods and no gap closure. The thermal performance of the two rods is very similar and the neutron absorption penalty due to stainless steel use could be compensating by combining small increase in U-235 enrichment and pitch size changes.展开更多
文摘The welding joints of Zircaloy 4 plates obtained by diffusion welding at 800°C under pressure in vacuum were cracked during autoclave tests at 400°C superheated steam after exposure longer than 150 days. The section of specimens was examined by optical microscopy and the composition at the tips of cracking was analyzed by electron microprobe. The result shows that the combination of oxidation and hydriding induced cracking is responsible for this failure of the welding joints.
文摘In order to investigate the effect of lanthanum ion implantation on theoxidation behavior of zircaloy at 500℃, Zircaloy specimens were implanted by lanthanum ions with adose range from 5xl0^(16) to 2xl0^(17) ions/cm^2 at room temperature, and then oxidized at 500℃ for100 min. The valence of the oxides in the scale was analyzed by X-ray Photoelectron Spectroscopy(XPS). The phase structures of the oxides in the scale were examined by Glancing Angle X-rayDiffraction (GAXRD). With the increase of implanted lanthanum ions dose, the phase structures in theoxide scale are transformed from monoclinic zirconia to hexagonal one and then to monoclinic oneagain. The measurement of weight gain showed that a similar change from the decreased gain toincreased one again is achieved in the oxidation behavior of lanthanum ion implanted zircaloycompared with that of as-received zircaloy.
基金Project supported by the National Basic Research Program of China(Grant No.2011CB610501)the Funds from the State Key Laboratory of Surface and Chemistry,China(Grant No.SPC 201102)the Reactor Fuel and Materials Laboratory,China(Grant No.STRFML-2013-05)
文摘In this work, hydrogen absorption and the permeation behavior of the passive layer formed on zircaloy-4 are in- vestigated. Potentiodynamic polarization, Mott-Schottky analysis, electrochemical impedance spectroscopy, and Raman scattering spectroscopy are employed to characterize the passive defects before and after hydrogen permeation. It is found that the nanoscale passive ZrO2 films play an important role in the resistance against corrosion; hydrogen impingement, however, reduces the passive impedance towards hydrothermal oxidation. The increase of defects （vacancies） in passive film is probably attributed to the degradation. We believe that this finding will provide valuable insight into the understanding of the corrosion mechanism of zircaloys used in light water reactors.
文摘Fatigue dislocation configurations of Zircaloy-4 at 470℃×1h stress-relieved condition and 620℃×1h recrystallized condition were analyzed using TEM. Theresults show that: {1 0 1 0} prismatic slip is the primary deformation mode at RT. Prismatic and pyramidal slips are activated simultaneously at 400℃. The typicalsubstructure is the elongated dislocation lines at RT; whereas at 400℃, it is rectangularcells in stress-relieved specimens, and elongated cells plus dipole perpendicular cellboundary in recrystallized specimens. The relationship map among dislocation configuration, test temperature and cyclic strain range is established, finally.
基金Supported by the National Natural Science Foundation of China (Grant No. 50461001)Guangxi Science and Technology Fund (Grant Nos. 0575-18, 0639003)Science Fund of Guangxi University (Grant No. 2005ZD04)
文摘The contrastive corrosion experiments between surface nanocrystallined Zircaloy-4 and coarse-grained Zircaloy-4 under the condition of 673 K/10.3 MPa in pure water are carried out, and the microstructure of oxide films has been studied. The results indicate that the growth rate of oxide films formed on the nanocrystalline Zircaloy-4 is lower than that of oxide films formed on the coarse-grained Zircaloy-4. Simultaneously, the oxide/metal interface of the former is more regular and glossy than that of the latter. For nanocrystalline Zircaloy-4, the low oxygen diffusion rate through the oxide/metal interface can hinder the reaction of oxygen ion with metal ion. Furthermore, more tetragonal ZrO2 are observed in the oxide films, which can delay the martensite phase transition from tetragonal to monoclinic phase in oxide films.
基金This work was financially supported by the National Science Foundation of China (No.50571056)the National Science Foundation of Shanghai (No.04ZR14057).
文摘After being treated in different ways, Zr-Sn-Nb-Fe alloy specimens are exposed in 0.01mol/L LiOH aqueous solution at 350℃ under 16.8 MPa. The examination of microstructures and second phase particles （SPPs） of these specimens was carried out by high-resolution transmission electron microscopy （HR-TEM）. The specimens treated at 800℃ before the final cold rolling have a better corrosion resistance than those treated at 680℃, and the specimens treated at 500℃, after the final cold rolling, have a better corrosion resistance than those treated at 560℃. TEM examination shows that the SPPs existing in the 800℃/500℃ specimen, which has the best corrosion resistance, contains a lot of Nb element, which results in the reduction of the niobium content in the α-Zr solid solution.
文摘In the past, stainless steel was utilized as cladding in many PWRs （pressurized water reactors）, and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as cladding material mainly due to its lower neutron absorption cross section. Now, stainless steel cladding appears as a possible solution for safety problems related to hydrogen production and explosion as occurred in Fukushima Daiichi accident. The aim of this paper is to discuss the steady-state irradiation performance using stainless steel as cladding. The results show that stainless steel rods display higher fuel temperatures and wider pellet-cladding gaps than Zircaloy rods and no gap closure. The thermal performance of the two rods is very similar and the neutron absorption penalty due to stainless steel use could be compensating by combining small increase in U-235 enrichment and pitch size changes.
