ISSN 0412-1961
CN 21-1139/TG
Started in 1956

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    RESEARCH TRENDS ON MICRO AND NANO--SCALE MATERIALS DEGRADATION IN NUCLEAR POWER PLANT
    HAN En--Hou
    Acta Metall Sin, 2011, 47 (7): 769-776.  DOI: 10.3724/SP.J.1037.2011.00441
    Abstract   PDF (2328KB) ( 1355 )
    The recent research status of materials degradation in nuclear power plant (NPP) has been analyzed. The main research progresses include: kinetics of corrosion electrochemistry in high temperature pressurised water, preferential intergranular oxidation and the degradation of grain boundary strength, special grain boundary optimized materials corrosion resistance, tight crack of stress corrosion cracking, nano-scale atom cluster formation and effects. The research trends and key problems have been proposed, such as kinetics of corrosion electrochemistry in high temperature high pressure water, especially the effect of contaminating species on micro-process of corrosion; characterization of microstructure, physical, chemical and mechanical properties of surface film and near-surface materials, repassivation behavior of passive film, especially ion diffusion micro-process in the surface film and near-surface materials; the effects of as-received surface and water chemistry on crack incubation and initiation; application of materials degradation mechanisms. In order to characterize the material environmental behavior, it is important to control the research and environment condition,  and to develop new test methods in simulated NPP environment.
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    RADIATION-NDUCED EMBRITTLEMENT AND LIFE EVALUATION OF REACTOR PRESSURE VESSELS
    LU Zheng
    Acta Metall Sin, 2011, 47 (7): 777-783.  DOI: 10.3724/SP.J.1037.2011.00265
    Abstract   PDF (925KB) ( 1162 )
    Neutron irradiation can lead to the embrittlement and the shift of ductile-brittle transition temperature (DBTT) of the reactor pressure vessel (RPV) steels, which is the potential threat to the safe operation of reactors. Surveillance program for RPVs, irradiation damage and embrittlement of RPV steels and safe evaluation of RPVs were reviewed in this paper.
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    ANALYSIS AND VALIDATION OF PIPE WALL THINNING DUE TO FLOW ACCELERATED CORROSION
    Masanori Naitoh CHEN Yaodong Shunsuke Uchida Hidetoshi Okada
    Acta Metall Sin, 2011, 47 (7): 784-789.  DOI: 10.3724/SP.J.1037.2011.00312
    Abstract   PDF (2222KB) ( 1175 )
    Flow accelerated corrosion (FAC) is one of phenomena which are challenging safety operation of power plant. The mechanism and dominant factor contributing to its ocurrence are illustrated. In parrallel, a dedicated FAC simulation code package, DRAWTHREE, its physical models and structure, as well as methodology and procedure for FAC and wall thinning evaluation are introduced. The code is then applied to the simulation of FAC and prediction of wall thinning rate, and the simulated results agree well with experimental and plant measured data. Finally, some countermeasures against FAC for different types of power plant are proposed.
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    CORROSION FATIGUE OF NUCLEAR--GRADE STAINLESS STEEL IN
    HIGH TEMPERATURE WATER AND ITS ENVIRONMENTAL FATIGUE DESIGN MODEL
    WU Xinqiang XU Song HAN En-Hou KE Wei
    Acta Metall Sin, 2011, 47 (7): 790-796.  DOI: 10.3724/SP.J.1037.2011.00162
    Abstract   PDF (1482KB) ( 1230 )
    The high safety of light water reactor nuclear power plants (NPPs) requires very strict design standard and service property of pressure boundary components materials. The service degradation and life assessment of the components materials primarily depend on the understanding of environmentally assisted failure mechanism, accumulation of service property data and construction of evaluation models. Currently domestic NPPs are relying on foreign design, operation and life assessment standards. However, recent experimental data indicate that even the ASME design fatigue code may be deficient in safety margin under certain conditions of loading and environment. In the present work, based on the corrosion fatigue tests in simulated NPPs' high temperature pressurized water, the corrosion fatigue behavior and environmentally assisted failure mechanism of domestic nuclear-grade stainless steel have been investigated. The factors affecting fatigue life of nuclear grade stainless steel in high temperature water were evaluated. A design fatigue model was constructed by taking environmental degradation effects into account and the corresponding design curves were given for the convenience of engineering applications.
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    STRESS CORROSION CRACKING BEHAVIOR OF DISSIMILAR METAL WELD A508/52M/316L IN HIGH TEMPERATURE WATER ENVIRONMENT
    LI Guangfu LI Guanjun FANG Kewei PENG Jun YANG Wu ZHANG Maolong SUN Zhiyuan
    Acta Metall Sin, 2011, 47 (7): 797-803.  DOI: 10.3724/SP.J.1037.2011.00316
    Abstract   PDF (1514KB) ( 1551 )
    The stress corrosion cracking (SCC) behavior of advanced dissimilar metal weld A508/52M/316L in simulated primary water environments of pressurized water reactor (PWR) at 290 ℃ was investigated by means of slow strain rate testing (SSRT). The tests were performed at various applied electrode potentials which correspond to the electrochemical conditions of the weld in various water environments, from low potentials with ideal water chemistry to high potentials with oxygen-contaminated water chemistry. The weld exhibits complicated microstructure and chemical composition distributions, especially, significant changes appear around the A508/52M interface and the 52M/316L interface. For tensile specimens in SSRT, sharp notches were machined at important and typical places, i.e., at the two interfaces and in the bulk parts of the low alloy steel, Ni base weld metal and stainless steel of the weld. Results showed that the specimens always failed in bulk zone of the Ni base weld metal with ductile appearances when tested in the potential range from -780 mV to -300 mV (vs SHE). When electrode potential was raised into the range from -200 mV to +200 mV which corresponds to oxygen-contaminated water chemistry, the weld exhibits significant SCC. The area around the A508/52M interface is the weakest place, transgranular stress corrosion cracking (TGSCC) happened both along the interface and in A508 heat affected zone (HAZ), intergranular stress corrosion cracking (IGSCC) occurred in the Ni base weld metal close to the interface. The cracking mechanism and the engineering practical significance were discussed.
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    DEPOSITION MECHANISM OF PARTICLE-LIKE CORROSION PRODUCT IN TURBULENT DUCT
    YAO Jun Michael Fairweather LI Ning
    Acta Metall Sin, 2011, 47 (7): 804-808.  DOI: 10.3724/SP.J.1037.2011.00323
    Abstract   PDF (1025KB) ( 994 )
    The deposition of corrosion products (particles) in cooling circuits of water-cooled nuclear reactors has been investigated using large eddy simulation and Lagrangian method (Reynolds number 2.5×105). A particle equation of motion including Stokes drag, lift, buoyancy and gravitational forces is used for particle trajectory analysis. The fluid-particle effect is considered and the particle-particle interact is ignored in this work. Results obtained from the fluid field calculation showed good agreement with experimental data and the predictions of direct numerical simulations. The particle size, drag force, shear-induced lift force and gravity all affected the particle deposition process. The small sized particles tended to deposite near the duct center while large sized particles tended to deposite near the duct edge, which become more obvious with increasing particle size. Close to the bottom of the duct, the particle number density increased with particle size increasing, and a high concentration of large particles appeared in the region with flow velocities lower than the mean, while small particles distribute evenly throughout the flow.
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    EFFECTS OF Cl- CONCENTRATION AND TEMPERATURE ON THE CORROSION BEHAVIOR OF ALLOY 690 IN BORATE BUFFER SOLUTION
    HUANG Fa WANG Jianqiu HAN En-Hou KE Wei
    Acta Metall Sin, 2011, 47 (7): 809-815.  DOI: 10.3724/SP.J.1037.2011.00212
    Abstract   PDF (1426KB) ( 1235 )
    The effects of Cl$^{-}$ concentration (0.5-2 mol/L) and temperature (25-80 ℃) on the corrosion behavior of alloy 690 in borate buffer solution were investigated using potentiodynamic polarization (PD), electrochemical impedence spectroscopy (EIS) and semiconductor capacitance method (Mott-Scottky relation). Atomic force microscope (AFM), X-ray photoelectron spectroscopy (XPS) and potential--pH diagrams were employed to analyze the corrosion products. All of the polarization curves exhibited two passive regions and intergranular corrosion was observed on all samples. With the increase in both Cl- concentration and temperature, the corrosion potential decreased and the corrosion current density became larger. Furthermore, increasing the temperature also resulted in lower pitting potentials and narrower passive regions. After anodic polarization for 45 min, the film formed in the first passive region with lower potential was composed of Cr2O3, Fe2O3 and Ni(OH)2, and behaved like a mixed-type semiconductor, while a thicker but less compact Ni2O3 film was formed in the second passive range with higher potential and behaved like a n-type semiconductor. The influences of Cl- and temperature on the corrosion behavior were discussed.
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    EFFECTS OF NITROGEN CONTENT ON MICROSTRUCTURE AND MECHANICAL PROPERTY OF ALLOY 690
    LI Shuo CHEN Bo MA Yingche GAO Ming LIU Kui
    Acta Metall Sin, 2011, 47 (7): 816-822.  DOI: 10.3724/SP.J.1037.2011.00158
    Abstract   PDF (1292KB) ( 1180 )
    The microstructure and mechanical property at room temperature of Ni-30Cr-10Fe-xN (x=0.001, 0.011, 0.018, 0.030, mass fraction, %) base alloy 690 were investigated. The OM observation shows that the amounts of precipitated nitrides and the annealing twins in the matrix increase with more nitrogen added in the alloy. The carbide precipitation along boundary and the chromium depletion near the grain boundary are modified by nitrogen addition, carbide becomes less and chromium depletion is mitigated. The tensile test result shows that both of the yield and tensile strengths are increased by about 50 MPa when nitrogen content increased from 0.001% to 0.030%, while the ductility lowers a little. The fractographs exhibit typical ductile dimple pattern and the sizes of the dimples reduce with increasing nitrogen content in the alloy.
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    ANALYSES OF SURFACE OXIDE FILMS ON GROUND ALLOY 690TT AFTER IMMERSION FOR DIFFERENT TIMES
    ZHANG Zhiming WANG Jianqiu HAN En-Hou KE Wei
    Acta Metall Sin, 2011, 47 (7): 823-830.  DOI: 10.3724/SP.J.1037.2011.00206
    Abstract   PDF (1610KB) ( 925 )
    The morphologies and structures of surface oxide films grown on ground Ni base alloy 690TT after immersion in the simulated hydrogenated primary water of pressured water reactors (PWRs) for different times were analyzed by various methods. After immersion for 35 h,\linebreak the ground alloy 690TT was covered with compact oxide particles. With the increase of the immersion time, the sample surfaces were covered with scattered big oxide particles and compact small oxide particles. Regardless of the immersion time, the grown oxide films are composed of spinel oxides and metallic Ni. After immersion for 720, 1440 and 2160 h, the oxide films are composed of three layers: the outmost layer is the separated big oxide particles which are rich in Fe and Ni; the intermediate layer is the compact small oxide particles rich in Cr, Fe and Ni; the inner layer is the continuous Cr oxides. The peak decompositions of the XPS results revealed that the Cr oxides in the inner layer are probably Cr2O3. The intermediate and inner layers in the oxide films could restrain the outward diffusion of metal atoms and also the inward diffusion of the oxygen atoms and then protect the matrix from further corrosion well. The average corrosion rate of the intermediate and inner layer decreased gradually with the immersion time increasing. Grinding treatment accelerated the growth of protective oxide film on alloy 690TT in the studied solution.
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    ANALYSES OF SURFACE OXIDE FILMS ON ELECTROPOLISHED ALLOY 690TT AFTER IMMERSION FOR DIFFERENT TIMES
    ZHANG Zhiming WANG Jianqiu HAN En-Hou KE Wei
    Acta Metall Sin, 2011, 47 (7): 831-838.  DOI: 10.3724/SP.J.1037.2011.00300
    Abstract   PDF (1624KB) ( 925 )
    The corrosion of Ni base alloys in high temperature and high pressure water is affected by samples surface statuses. The morphologies and structures of surface oxide films grown on electropolished (EP) alloy 690TT after immersion in the simulated hydrogenated primary water of pressured water reactors (PWRs) for different times were analyzed by AFM, SEM, TEM, EDS and XPS. After immersion for 15 and 35 h, the EP alloy 690TT samples were covered with columnar oxides. With the increase of the immersion time, the sample surfaces were covered with scattered big oxide particles and loose needle-like oxides. Regardless of the immersion time, the formed oxide films are composed of spinel oxides and metallic Ni. After immersion for 720, 1440 and 2160 h, the oxide films are composed of three layers: the outmost layer is the separated big oxide particles which are rich in Fe and Ni; the intermediate layer is the loose needle--like oxides rich in Ni; the inner layer is the continuous and compact Cr oxides. The peak decomposition of the XPS revealed that the Cr oxides in the inner layer are probably Cr2O3. Only the inner layer in the oxide film could restrain the outward diffusion of metal atoms and also the inward diffusion of the oxygen atoms and then played the role of barrier layer well. Electropolishing treatment disadvantaged the fast growth of protective oxide film on alloy 690TT in the studied solution. The average corrosion rate of the inner layer does not decrease gradually with increasing the immersion time. After immersion for 2160 h, the oxide film still could not protect the matrix from further corrosion.
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    MICROSTRUCTURE NEAR SCRATCH ON ALLOY 690TT AND STRESS CORROSION INDUCED BY SCRATCHING
    MENG Fanjiang WANG Jianqiu HAN En-Hou SHOJI Testuo KE Wei
    Acta Metall Sin, 2011, 47 (7): 839-846.  DOI: 10.3724/SP.J.1037.2011.00213
    Abstract   PDF (1761KB) ( 1137 )
    The microstructure and stress corrosion cracking (SCC) behavior of scratched zone on alloy 690TT were studied by using microhardness, TEM, EBSD-OIM and immersion experiment in caustic solution. It was found that a deformed hardening layer with a dimension range of 100 $\mu$m was produced near the scratch. TEM and EBSD-OIM observations showed that the grains at shallow surface of scratch groove were refined to nano-size. SCC tests for scratched alloy 690TT were performed in caustic solution at high temperature with or without addition of lead oxides. The results showed that SCC cracks initiated and propagated at scratch banks and scratch grooves. Grain boundaries, twin boundaries deformed and microcracks produced during scratching process are preferential sites for SCC. The oxide films formed on scratch groove were loosed by lead. The SCC crack length increased with increase of lead content. Scratched alloy 690TT is susceptible to SCC.
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    INFLUENCES OF TiN INCLUSION ON CORROSION AND STRESS CORROSION BEHAVIORS OF ALLOY 690 TUBE IN HIGH TEMPERATURE AND HIGH PRESSURE WATER
    LI Xiaohui HUANG Fa WANG Jianqiu HAN En-Hou KE Wei
    Acta Metall Sin, 2011, 47 (7): 847-852.  DOI: 10.3724/SP.J.1037.2011.00205
    Abstract   PDF (1266KB) ( 1158 )
    TEM and SEM were used for studying the existing forms and distribution states of TiN inclusions in alloy 690. The corrosion and stress corrosion behaviors of alloy 690 were investigated through high temperature high pressure electrochemical measurement and immersion tests. TiN particles are the main inclusion in alloy 690 and randomly distribute in austenitic matrix. Corrosion experiments showed that pitting corrosion occurs preferentially at the inclusions containing Ti with high potential in simulated primary water of pressurized water reactor (PWR). In high temperature high pressure alkali solution containing lead, the region near TiN inclusion and close to the side of matrix is the preferred position for corrosion, and TiN inclusions at grain boundary can subsequently cause local stress concentration on grain boundary, both of which lead to intergranular stress corrosion cracking in alloy 690.
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    STUDY OF CARBIDE PRECIPITATION AT GRAIN BOUNDARY IN NICKEL BASE ALLOY 690
    LI Hui XIA Shuang ZHOU Bangxin PENG Jianchao
    Acta Metall Sin, 2011, 47 (7): 853-858.  DOI: 10.3724/SP.J.1037.2011.00197
    Abstract   PDF (1084KB) ( 1164 )
    The morphology and orientation relationship between carbide precipitated at grain boundary and both side matrixes in nickel base alloy 690 aged at 715 ℃ for 2-200 h after solution treated were investigated by HRTEM, SEM and EBSD. The results show that the boundary carbide is easy to nucleate in the grain with high indexed crystal plane parallel to boundary, and the carbide has coherent orientation relationship (COR) with this grain. The carbide grows preferentially into the grain without COR between carbide and grain, which leads to lower chromium concentration in this side matrix near boundary. The asymmetry of chromium concentration profile leads to the different corrosion resistance of the two side grains nearby the grain boundary.
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    EFFECT OF ORIGINAL GRAIN SIZE ON THE BOUNDARY NETWORK IN ALLOY 690 TREATED BY GRAIN BOUNDARY ENGINEERING
    LIU Tingguang XIA Shuang LI Hui ZHOU Bangxin CHEN Wenjue
    Acta Metall Sin, 2011, 47 (7): 859-864.  DOI: 10.3724/SP.J.1037.2011.00196
    Abstract   PDF (1300KB) ( 991 )
    Effect of original grain size on the grain boundary character distribution (GBCD) in alloy 690 after treatment by grain boundary engineering (GBE) was studied using electron backscatter diffraction (EBSD) and orientation image microscopy (OIM). The proportion of low-ΣCSL grain boundary and grain boundary network after GBE are obviously influenced by the original grain size. Optimized GBE grain boundary network after the same annealing process can be obtained by altering the cold work degree based on the original grain size. Thus the microstructure after GBE is influenced by the combined effect of original grain size and deformation amount before annealing. Mean strain of grain is used to express the combined effect. A proper value of mean strain of grain is necessary to achieve the desired GBE grain boundary network.
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    EFFECT OF Nb ON THE CORROSION RESISTANCE OF Zr-4 ALLOY IN SUPERHEATED STEAM AT 500 ℃
    YAO Meiyi LI Shilu ZHANG Xin PENG Jianchao ZHOU Bangxin ZHAO Xushan SHEN Jianyun
    Acta Metall Sin, 2011, 47 (7): 865-871.  DOI: 10.3724/SP.J.1037.2011.00106
    Abstract   PDF (1073KB) ( 994 )
    The corrosion resistance of Zr-4+xNb alloys (x=0.1-0.3, mass fraction, %) was investigated in a superheated steam at 500 ℃ and 10.3 MPa by autoclave tests. The microstructure of the alloys and fracture surface morphology of the oxide film formed on the alloys were observed by TEM and SEM, respectively. Results show that no nodular corrosion appears on Zr-4+xNb alloys in 500 ℃/10.3 MPa superheated steam even for 500 h, which is related to the higher Nb concentration dissolved in α-Zr matrix. The Nb dissolved in α-Zr matrix can restrain the nucleation of nodular corrosion, thus improve the nodular corrosion resistance. The uniform corrosion resistance of Z-4+xNb alloys is lowered with the increase of Nb content, which is related to the decrease of the solid solution concentration of (Fe+Cr) in α-Zr matrix and the precipitation of the second phase particles of Zr(Fe, Cr, Nb)2, and the two aspects will accelerate the microstructural evolution of the oxide film during corrosion process to promote the formation of pores and micro-cracks.
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    EFFECT OF Cu CONTENT ON THE CORROSION RESISTANCE OF Zr-0.80Sn-0.34Nb-0.39Fe-0.10Cr-xCu ALLOY IN SUPERHEATED STEAM AT 500 ℃
    YAO Meiyi ZHANG Yu LI Shilu ZHANG Xin ZHOU Jun ZHOU Bangxin
    Acta Metall Sin, 2011, 47 (7): 872-876.  DOI: 10.3724/SP.J.1037.2011.00236
    Abstract   PDF (860KB) ( 881 )
    The effect of Cu content on the corrosion resistance of Zr-0.8Sn-0.34Nb-0.39Fe-0.1Cr-xCu alloys (x=0.05-0.5, mass fraction, %) was investigated in superheated steam at 500 ℃ and10.3 MPa by autoclave tests. The microstructures of the alloys are observed by TEM. The results show that (0.05-0.5)Cu addition has little effect on the corrosion resistance of the alloys. When x is below 0.2, the precipitates Zr(Fe, Cr, Nb)2 with hcp structure and Zr3Fe containing Cu with orthorhombic structure are detected. When x is above 0.2, besides Zr(Fe, Cr, Nb)2 and Zr3Fe containing Cu, the precipitate of Zr2Cu with tetragonal structure is also detected. Zr(Fe, Cr, Nb)2 precipitates are smaller than the precipitates containing Cu in size. The precipitates containing Cu are found in the alloy even with 0.05Cu, indicating that the Cu content in α-Zr matrix is very small. Therefore, the reason that the Cu content has little effect on the corrosion resistance of the alloys is maybe related to the lower Cu content in α-Zr matrix.
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    EFFECT OF Cu-ADDITION ON THE CORROSION BEHAVIOR OF Zr-2.5Nb ALLOYS IN 500 ℃/10.3 MPa SUPERHEATED STEAM
    LI Qiang LIANG Xue PENG Jianchao YU Kang YAO Meiyi ZHOU Bangxin
    Acta Metall Sin, 2011, 47 (7): 877-881.  DOI: 10.3724/SP.J.1037.2011.00275
    Abstract   PDF (990KB) ( 1471 )
    Zr-2.5Nb-xCu (x=0.2, 0.5, mass fraction, %) specimens were annealed at 580 ℃ for 50 h and 620 ℃ for 2 h after β-quenching and cold rolling, and then corroded in autoclave with 500 ℃/10.3 MPa superheated steam. SEM and TEM equipped with EDS were employed to investigate their corrosion behaviors through analyzing the microstructures of the fracture surface of the oxide layers or of the cross-section of oxide layers. It was found that the addition of copper can improve the corrosion resistance of Zr-2.5Nb alloy, which is closely related to the increases in the fraction and size of columnar grains in the oxide film.
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    CORROSION BEHAVIOR OF Zr(Fex, Cr1-x)2 ALLOYS IN 400℃ SUPERHEATED STEAM
    CAO Xiaoxiao YAO Meiyi PENG Jianchao ZHOU Bangxin
    Acta Metall Sin, 2011, 47 (7): 882-886.  DOI: 10.3724/SP.J.1037.2011.00251
    Abstract   PDF (1164KB) ( 1016 )
    To study the corrosion behavior of second phase particles in zirconium alloys, Zr(Fex, Cr1-x)2 (x=1, 2/3, 1/3) metallic compounds which have the same composition as the second phase particles in Zr-4 alloy were prepared by vacuum non-consumable arc melting. XRD and energy filtered TEM were employed for analyzing the corrosion products, the element distribution and grain morphology after corrosion tests of Zr(Fex, Cr1-x)2 metallic compounds powder at 400 ℃ and 10.3 MPa superheated steam with different exposure times. The results show that Cr has a very strong effect on the corrosion resistance of Zr(Fex, Cr1-x)2 metallic compounds, increasing Cr content can improve the corrosion resistance. When Zr(Fex, Cr1-x)2 oxidation starts, zirconium oxide is formed while elements Fe and Cr are expelled from the zirconium oxide due to their low solid solubility in the oxide. α-Fe(Cr) and γ-Fe(Cr) are formed and then oxidized to the stable corrosion product (Fe, Cr)3O4. The different corrosion behaviors of metallic compounds will affect the microstructure evolution of zirconium oxide layer differently during the corrosion process, and hence affect the corrosion resistance of zirconium alloys.
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    INVESTIGATION OF MICROSTRUCTURE OF OXIDE LAYERS FORMED INITIALLY ON Zr-4 ALLOY
    DU Chenxi PENG Jianchao LI Hui ZHOU Bangxin
    Acta Metall Sin, 2011, 47 (7): 887-892.  DOI: 10.3724/SP.J.1037.2011.00176
    Abstract   PDF (1135KB) ( 917 )
    Zr--4 specimens with coarse grain of 0.2-0.8 mm were prepared to investigate the anisotropic growth of oxide layers formed initially on the grain surface with different orientations during corrosion tests in autoclave at 360 ℃/18.6 MPa in 0.01 mol/L LiOH aqueous after 5 h exposure. SEM, EBSD and HRTEM were adopt to measure the thickness of oxide layers, to determine the grain orientation of the matrix surface and to investigate the microstructure of oxide layers. The thicknesses of oxide layers formed on different grains varied in the range of 376-455 nm. The thickest oxide layers were detected on the grains with the orientations nearby basal plane (0001) and prismatic plane (0110). The oxide layers have monoclinic, cubic, tetragonal crystal structures. Besides the thickness difference of oxide layers, the crystal structure and misorientation of nano-grains in oxide layers formed on different grains were also significantly different, and the most complicated oxide layer was formed on the grain with orientation nearby (0001) plane. Such kind of microstructure has more crystal defects, and larger ability for promoting the diffusion of oxygen ions and the growth of oxide layer.
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    OXIDATION BEHAVIOR OF THE β-Nb PHASE PRECIPITATED IN Zr-2.5Nb ALLOY
    LI Qiang LIANG Xue PENG Jianchao LIU Renduo YU Kang ZHOU Bangxin
    Acta Metall Sin, 2011, 47 (7): 893-898.  DOI: 10.3724/SP.J.1037.2011.00276
    Abstract   PDF (1301KB) ( 1179 )
    The corrosion behavior for Zr--2.5Nb specimens heat--treated at 580 ℃ for 50 h after $\beta$--quenching
    and cold rolling has been investigated in 500 ℃/10.3 MPa superheated steam by autoclave tests. HRTEM
    equipped with EDS was employed to investigate the matrix microstructure and the oxidation behavior of the
    $\beta$--Nb second phase particles (SPPs). It was found that many $\beta$--Nb SPPs with small sizes ($<$100 nm)
    randomly precipitated after heat treating at 580 ℃ for 50 h. It was noted that the $\beta$--Nb SPPs were more
    slowly oxidized than the zirconium matrix. The $\beta$--Nb SPPs of bcc structure were oxidized to form the mixed
    structure of amorphous oxide and crystalline NbO$_{2}$ at the initial oxidation stage, and then the amorphous phase was changed to the
    main structure at the middle oxidation stage, finally the niobium oxides were dissolved into the corrosion medium.
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    EFFECT OF β-QUENCHING ON THE CORROSION RESISTANCE OF Zr-4 ALLOY IN LiOH AQUEOUS SOLUTION
    SHEN Yuefeng YAO Meiyi ZHANG Xin LI Qiang ZHOU Bangxin ZHAO Wenjin
    Acta Metall Sin, 2011, 47 (7): 899-904.  DOI: 10.3724/SP.J.1037.2011.00233
    Abstract   PDF (1220KB) ( 1205 )
    To investigate the effect of $\beta$--quenching on the corrosion behavior, Zr--4 specimens were
    $\beta$--quenched and subsequently annealed at 480---600 ℃ for 2---200 h. The corrosion tests were carried out in
    0.01 mol/L LiOH aqueous solution at 360 ℃/18.6 MPa. The microstructure of Zr--4 specimens and fracture
    surface morphology of oxide films on the corroded specimens were observed by TEM and HRSEM,
    respectively. The results show that Zr--4 specimens exhibit excellent corrosion resistance as Zr--Sn--Nb
    alloys in LiOH aqueous solution after increasing the solid solution content of Fe and Cr in $\alpha$--Zr matrix by
    suitable quenching rate from $\beta$--phase to avoid the formation of $\beta$--Zr.
    When the cooling rate is too fast, the specimens show worse corrosion resistance due to the formation
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    INVESTIGATION ON THE PRECIPITATION AND STRUCTURAL EVOLUTION OF Cu-RICH NANOPHASE IN RPV MODEL STEEL
    XU Gang CHU Dafeng CAI Linling ZHOU Bangxin WANG Wei PENG Jianchao
    Acta Metall Sin, 2011, 47 (7): 905-911.  DOI: 10.3724/SP.J.1037.2011.00178
    Abstract   PDF (1350KB) ( 987 )
    The crystal structure of Cu-rich nanophase in reactor pressure vessel (RPV) model steel was investigated by means of HRTEM, EDS and APT methods. RPV model steel was prepared by vacuum induction furnace melting with higher content of Cu (0.6%, mass fraction). The ingot about 40 kg in weight was forged and hot rolled to 4 mm in thickness and then cut to specimens of 40 mm×30 mm. Those specimens were further heat treated by 880 ℃ for 0.5 h water quenching and 660 ℃ for 10 h tempering, and finally aged at 370 ℃ for 6000 h. It was observed that the Cu atoms segregated on {110} planes of α-Fe matrix in a period every three layers and the distortion of crystal lattice was induced by inner stress produced by the segregation of Cu atoms during the nucleation of the precipitation of Cu-rich nanophase. The inner stress in Cu-rich regions enhanced with the increase of Cu concentration as well as the enlargement in size. It was also observed that the Cu-rich regions underwent a transformation from bcc structure to 9R structure with twins by means of a shear along the {110} plane of α-Fe matrix. Finally, the Cu-rich clusters transformed to fcc  structure with further increase of Cu content. The results obtained by APT analysis show that the equivalent diameter and the number density of Cu-rich clusters are about 1-8 nm and 0.71×1023 m-3, respectively. The content of 3%-8% Ni and Mn within the Cu-rich clusters was detected. It is suggested that the growth of Cu-rich clusters was restrained by the segregation of Ni and Mn atoms on the boundaries around the Cu-rich clusters.
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    MICROSTRUCTURE OF ANNEALED 12Cr13 STAINLESS STEEL AND ITS EFFECT ON THE IMPACT TOUGHNESS
    HAO Xianchao GAO Ming ZHANG Long ZHAO Xiujuan LIU Kui
    Acta Metall Sin, 2011, 47 (7): 912-916.  DOI: 10.3724/SP.J.1037.2011.00159
    Abstract   PDF (1122KB) ( 1092 )
    The microstructure and its effect on impact toughness of the annealed 12Cr13 stainless steel were investigated by OM and SEM. The results show that the morphology of carbides is the main factor that influences the impact properties. Globular-like, block or strip-like carbides precipitated continuously during annealing. Fine globular-like carbides in the pre-existed martensite grains improved the impact toughness. Block or strip-like carbides at the grain boundaries or in the $\delta$--ferrite grains are detrimental to the impact property. The annealing temperature has significant effect on the carbide precipitation behavior. With the annealing temperature increasing from 760 ℃ to 860 ℃, carbides coarsen resulting in the decrease of the impact energy from 151 J to 106 J. The sample with only block or strip-like carbides has low impact energy of 5 J.
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    CHANGES OF MICROSTRUCTURE AND MECHANICAL PROPERTY OF THE CLAM STEEL AFTER LONG TERM AGING AT 600 ℃
    YANG Chunguang YAN Wei WANG Wei SHAN Yiyin YANG Ke WU Yican
    Acta Metall Sin, 2011, 47 (7): 917-920.  DOI: 10.3724/SP.J.1037.2011.00156
    Abstract   PDF (806KB) ( 933 )
    China low activation martensitic (CLAM) steel is a new reduced activation ferritic/martensitic steel developed in China to use for blanket/first-wall structures of the D-T fusion reactor. The microstructures and mechanical properties of the CLAM steel after aging at 600 ℃ for 1100  and 3000 h were investigated by SEM and TEM. The results showed that the prior austenite grain size did not increase with increasing aging time, while the precipitates at grain boundaries and inside grains were increased. The strength of CLAM steel was slightly increased after aging for 1100 h. When the aging time was prolonged to 3000 h, the strength of CLAM steel was decreased due to coarsening of precipitates. The ductile-brittle transition temperature (DBTT) was increased after aging for 1100 h. However, when the aging time increased to 3000 h, the DBTT lowered to the as heat-treated level. The changes of strength, toughness and DBTT after aging were interpreted from the view points of precipitation and coarsening behaviors of precipitates. It was also noticed that the low W content in the CLAM steel can effectively postpone the formation of Laves phase.
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    INVESTIGATION ON HEAT TREATMENT OF A DUPLEX STAINLESS STEEL FOR NUCLEAR POWER PLANT IMPELLER
    LIANG Tian KANG Xiuhong HU Xiaoqiang LI Dianzhong
    Acta Metall Sin, 2011, 47 (7): 921-926.  DOI: 10.3724/SP.J.1037.2011.00165
    Abstract   PDF (1243KB) ( 881 )
    Effects of heat treatment on microstructure and mechanical property of a cast duplex stainless steel used as making impeller of seawater circular pump in nuclear power plant were investigated. The results demonstrate that the impact energy and elongation
    decrease sharply when heat-treated time was longer than 5 min at 800-850 ℃ or when the cooling rate was less than 7.5 ℃/min. The σ phases have been identified through SEM and XRD. The weakness of the mechanical properties is attributed to the precipitation of σ phase. With increasing aging time, the σ phase coarsened and the nucleation position moved from the γ/δ interfaces into the whole ferrite phase.
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    EFFECT OF Ni2+ IN OCCLUDED VOLUME ON THE OXIDATION BEHAVIOR OF 304 STAINLESS STEEL IN HIGH TEMPERATURE WATER
    KUANG Wenjun WU Xinqiang HAN En-Hou
    Acta Metall Sin, 2011, 47 (7): 927-931.  DOI: 10.3724/SP.J.1037.2011.00161
    Abstract   PDF (990KB) ( 830 )
    In the occluded volume of high temperature pressurized water loop in nuclear power plants, poor-controlled water chemistry due to sluggish flowage could induce accelerated degradation of structural materials. In this paper, an occluded volume shaped nuclear-grade 304 stainless steel sample was immersed in oxygenated high temperature water containing a certain concentration of Ni2+. It was found that from the outer to the inner side of the sample, the proportion of spinel in the oxide film decreased while that of hematite increased and Ni content in the outer oxide film decreased. It was thought that the concentration of Ni2+ could influence the oxidation behavior of material significantly. The occluded volume could impede the inward diffusion of Ni2+, causing a concentration gradient of Ni2+ along depth of the volume and gradual change of oxide film characteristic.
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    CORROSION BEHAVIORS OF NUCLEAR-GRADE STAINLESS STEEL AND FERRITIC-MARTENSITIC STEEL IN SUPERCRITICAL WATER
    ZHONG Xiangyu WU Xinqiang HAN En-Hou
    Acta Metall Sin, 2011, 47 (7): 932-938.  DOI: 10.3724/SP.J.1037.2011.00163
    Abstract   PDF (1240KB) ( 877 )
    The corrosion behaviors of nuclear-grade 304 stainless steel (304SS) and ferritic-martensitic steel P92 exposed to 400℃/25 MPa supercritical water were investigated. The exposed specimens were characterized by weight gain measurement, XRD, Raman spectroscopy and SEM. It is found that both materials show general corrosion and exponential kinetics in mass gain, and the mass gain of 304SS is approximately an order of magnitude less than that of steel P92. The oxide film on 304SS is rather thin and composed of Cr2O3, α-Fe2O3, Fe3O4 and spinel, some nodules were observed on the surface. While the oxide film on steel P92 consists of α-Fe2O3, Fe3O4 and spinel, more α-Fe2O3 exist in the outer surface of the oxide film. The surface morphology of oxide film on steel P92 changes from dense particles to porous network structure with increasing exposure time, which may be relative to the dissolution of oxide.
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    EFFECT OF GRAIN BOUNDARY NETWORK ON THE INTERGRANULAR STRESS CORROSION CRACKING OF 304 STAINLESS STEEL
    HU Changliang XIA Shuang LI Hui LIU Tingguang ZHOU Bangxin CHEN Wenjue
    Acta Metall Sin, 2011, 47 (7): 939-945.  DOI: 10.3724/SP.J.1037.2011.00184
    Abstract   PDF (1414KB) ( 1321 )
    The grain boundary network in a 304 stainless steel can be controlled by grain boundary engineering (GBE). The total length proportion of Σ3n coincidence site lattice (CSL) boundaries was increased to more than 70%, and the large size highly twinned grain-cluster microstructure formed through the treatment of GBE. Stress corrosion cracking (SCC) susceptibility of 304 stainless steel was evaluated through C-ring specimen tests conducted in acidified boiling 20%NaCl solution. Based on the characterization by SEM, EBSD and OM, it was found that the large grain-clusters associated with many interconnected Σ3-Σ3-Σ9 and Σ3-Σ9-Σ27 triple junctions produced by GBE arrest the IGSCC cracks and improve the resistance to IGSCC.
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    STUDY ON VOID HEALING BEHAVIOR DURING FORGING PROCESS FOR 25Cr2Ni4MoV STEEL
    LI Shijian SUN Mingyue LIU Hongwei LI Dianzhong
    Acta Metall Sin, 2011, 47 (7): 946-953.  DOI: 10.3724/SP.J.1037.2011.00157
    Abstract   PDF (1531KB) ( 1247 )
    Based on the measured stress-strain curves and thermo-physical data of 25Cr2Ni4MoV steel, a FEM model of void closure in heavy forging process was established through ABAQUS software. The void closure behaviors under different hot plasticity conditions have been investigated. The FEM results showed three distinct stages during the void closure. The voids with the same position in a cylindrical specimen close up at a similar height reduction ratio (ΔH/H0) around 25\% at different temperatures, which indicates that the void closure is not sensitive to the deformation temperature. On basis of the FEM results, a physical-simulated experiment for void bonding process has been performed through compression tests on a hollow cylindrical specimens, with deformation temperatures ranging from 900 ℃ to 1200 ℃, ΔH/H0 from 25\% to 45\% and a constant strain rate 0.01 s-1. The experimental results have shown that the ΔH/H0 for the void completely bonded at 1200, 1100 and 1000℃ are all about 35%, but increases to 45\% when the deformation temperature decreased to 900℃, which confirms that void bonding is a diffusion controlled process. Through the experimental results, it can be further demonstrated the high temperature combined with severe deformation can enhance the ability of atoms transition and decrease the micro-gap between contacted surfaces. Finally, the influencing factors on the bonding efficiency were investigated through SEM observation on the fractured surface of the tensile specimen with internal closed void.
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    MOLECULAR DYNAMICS SIMULATION OF MATRIX RADIATION DAMAGE IN Fe-Cu ALLOY
    HE Xinfu YANG Peng YANG Wen
    Acta Metall Sin, 2011, 47 (7): 954-957.  DOI: 10.3724/SP.J.1037.2011.00207
    Abstract   PDF (801KB) ( 1016 )
    The reactor pressure vessel (RPV) is the highest priority key component in nuclear power plants and is considered irreplaceable. The embrittlement of RPV steels is in general thought to be caused primarily by the formation of Cu-enriched clusters, the formation of matrix damage features, and the segregation of P atoms at grain boundaries. The displacement cascades in Fe-0.05Cu and Fe-0.3Cu alloys were studied by molecular dynamics in the current work. It was found that substitutional Cu does not significantly affect the amount, annihilation and recombination of defect produced by displacement cascade but affects the vacancy migration energy significantly. The self-interstitial atoms (SIAs) clusters and vacancies clusters are formed during the displacement cascades and affected by irradiation temperature.
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    SURFACE MODIFICATION OF WC-Ni CEMENTED CARBIDE FOR SEALS BY HIGH-INTENSITY PULSED ION BEAM IRRADIATION
    ZHANG Fenggang ZHU Xiaopeng WANG Mingyang LEI Mingkai
    Acta Metall Sin, 2011, 47 (7): 958-964.  DOI: 10.3724/SP.J.1037.2011.00228
    Abstract   PDF (1285KB) ( 957 )
    The WC-Ni cemented carbide, as a promising seal component material in nuclear power plant, was treated by high-intensity pulsed ion beam (HIPIB) with ion energy of 300 keV, ion current density of 300 A/cm2, i.e., at a power density of 108 W/cm2 and at a pulse duration of 70 ns up to 10 shots. The phase composition, surface morphology and element distribution in the surface of WC-Ni cemented carbide before and after HIPIB irradiation were investigated by using XRD, SEM and EPMA, and its properties were characterized by microhardness measurement and block-on-ring wear testing. It is found that the phase transformation from hexagonal WC to cubic β-WC1-x underwent in the irradiated surface layer, and the amount of β-WC1-x phase increased with increasing shot number. The surface remelting and selective ablation of the nickel binder phase resulted in the formation of hilly topography with numerous protrusions on the irradiated surfaces, and the dimension of protrusions expanded under repetitive irradiation. As increasing the irradiation up to 10 shots, a network of hill-valleys was finally produced on the irradiated surfaces but the surface is smoothed and densified in a micro scale. A hardened depth of 160 μm was obtained, which is attributable to the strong stress wave induced during the irradiation. As a result, the wear resistance of WC-Ni cemented carbides is considerably improved by a factor of 3 along with a 38% reduction in the friction coefficient after 10-shot irradiation.
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    EFFECTS OF RADIATION AND He ON MICROSTRUCTURES OF LOW ACTIVE FERRITIC/MARTENSITIC STEEL F82H
    TONG Zhenfeng DAI Yong YANG Wen YANG Qifa
    Acta Metall Sin, 2011, 47 (7): 965-970.  DOI: 10.3724/SP.J.1037.2011.00208
    Abstract   PDF (1031KB) ( 912 )
    Low active ferritic/martensitic steel, F82H, has been developed as a candidate material for structural
    applications of fusion reactors because it has relatively low shifts in ductile-to-brittle transition temperature
    (DBTT) and excellent irradiation swell resistance. More works have been done in recent years on the
    microstructure and mechanical properties of F82H steel before and after irradiation, but most of the tested samples were
    irradiated at low temperature (<400 ℃). In this work, the microstructure of F82H steel irradiated in the Swiss
    spallation neutron source SINQ in a temperature range of 150-450 ℃ and a dose range of 6.1-20.2 dpa was
    studied. Defect clusters and He bubble were observed by TEM in the irradiated specimens.
    The results showed that there existed high density He bubbles with size of 1.6 nm under irradiation temperature higher than
    208℃, irradiation dose higher than 9.5 dpa and He concentration 680×10-6. The effects of
    irradiation dose, irradiation temperature and He concentration on microstructure of F82H steel were discussed.
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