M5和Zirlo合金高温水蒸气氧化行为研究
|
摘要
为了给实际应用中国产M5、Zirlo合金的安全和寿命提供理论依据.文中通过对国产M5和Zirlo合金在高温高压水蒸气下进行1 000h的氧化试验,分析氧化动力学,结合扫描电镜和能谱分析仪观察试样氧化产物的微观形貌和成分变化,研究其在高温高压水蒸气环境下的氧化行为.在室温下对不同氧化时长的M5和Zirlo合金进行环向拉伸试验,分析其力学性能以及观察断口形貌.试验结果表明:M5和Zirlo合金氧化动力学方程分别为w_(M5)=0.001t~(0.766 1),R=0.935 15以及w_(Zirlo)=0.000 58t~(0.399 3),R=0.808 34.Zirlo合金1 000h单位面积氧化增重为0.008 4mg·mm~(-2),远小于M5合金0.21mg·mm~(-2).M5合金随着氧化时长的增加试样表面被针状的氧化产物完全覆盖,Zirlo合金随着氧化时长增加针状氧化产物增加至1 000h未完全覆盖试样表面.M5合金表面O元素比例随氧化时长增加逐步增大,Zirlo合金表面O元素随着氧化时长增加变化不大.Zirlo合金抗氧化性能优于M5合金.M5在氧化1 000h内强度影响不大,Zirlo合金的强度随着时间的延长有所下降,M5和Zirlo合金的伸长率和断面收缩率随着氧化时间的延长均有所下降,Zirlo合金的伸长率和断面收缩率都小于M5合金,氧化后M5合金的塑性优于Zirlo合金.两者的拉伸断口均为微孔聚集型断裂,断裂类型为韧性断裂.M5合金的力学性能较Zirlo合金更稳定.
In order to provide theoretical basis for the safety of domestic M5 and Zirlo alloy in practical application.The oxidation behavior of M5 and Zirlo alloy which were exposed to high-temperature and high-pressure steam was investigated.The oxidation test for domestic M5 and Zirlo alloy was carried out in high-temperature and high-pressure steam for 1000 hours,after which the micro-morphology and compositional change of the samples were observed by scanning electron microscopy(SEM)and energy dispersive spectrometer(EDS).Besides,in order to analyze their mechanical properties and to observe their fracture morphology,the ring tensile test for M5 and Zirlo alloy with different oxidation time was carried out at room temperature.The experimental results showed that:the oxidation kinetics curves of M5 and Zirlo alloy follow the parabolic rate law(w_(M5)=0.001t~(0.766 1),R=0.935 15,w_(Zirlo)=0.000 58t~(0.399 3),R=0.808 34).The oxidation weight gain per unit area of the alloy Zirlo which is 0.008 4mg·mm~(-2) is far less than that of the M5 which is 0.21mg·mm~(-2) alloy.As for alloy M5,the homogeneous and dense oxidation products were produced on the sample surface with the increase of the oxidation time.The surface oxidation of the Zirlo alloy is inhomogeneous after 1 000hours' oxidation.In addition,the surface oxygen element ratio of the alloy M5 gradually increased with the increase of the oxidation time while the surface oxygen element ratio of the Zirlo alloy showed no obvious change with increasing the oxidation time.The oxidation resistance of the Zirlo alloy is superior to that of the alloy M5.What's more,the mechanical property of M5 kept almost constant within the oxidation time of 1 000 hours,but for Zirlo alloy,it decreased to some extent within the oxidation time of 1 000 hours.Also,both of their tensile fractures belong to the micro-void accumulation fracture and the fracture mode is characterized to be ductile fracture.In general,the mechanical property of alloy M5 is more stable than that of Zirlo alloy.
In order to provide theoretical basis for the safety of domestic M5 and Zirlo alloy in practical application.The oxidation behavior of M5 and Zirlo alloy which were exposed to high-temperature and high-pressure steam was investigated.The oxidation test for domestic M5 and Zirlo alloy was carried out in high-temperature and high-pressure steam for 1000 hours,after which the micro-morphology and compositional change of the samples were observed by scanning electron microscopy(SEM)and energy dispersive spectrometer(EDS).Besides,in order to analyze their mechanical properties and to observe their fracture morphology,the ring tensile test for M5 and Zirlo alloy with different oxidation time was carried out at room temperature.The experimental results showed that:the oxidation kinetics curves of M5 and Zirlo alloy follow the parabolic rate law(w_(M5)=0.001t~(0.766 1),R=0.935 15,w_(Zirlo)=0.000 58t~(0.399 3),R=0.808 34).The oxidation weight gain per unit area of the alloy Zirlo which is 0.008 4mg·mm~(-2) is far less than that of the M5 which is 0.21mg·mm~(-2) alloy.As for alloy M5,the homogeneous and dense oxidation products were produced on the sample surface with the increase of the oxidation time.The surface oxidation of the Zirlo alloy is inhomogeneous after 1 000hours' oxidation.In addition,the surface oxygen element ratio of the alloy M5 gradually increased with the increase of the oxidation time while the surface oxygen element ratio of the Zirlo alloy showed no obvious change with increasing the oxidation time.The oxidation resistance of the Zirlo alloy is superior to that of the alloy M5.What's more,the mechanical property of M5 kept almost constant within the oxidation time of 1 000 hours,but for Zirlo alloy,it decreased to some extent within the oxidation time of 1 000 hours.Also,both of their tensile fractures belong to the micro-void accumulation fracture and the fracture mode is characterized to be ductile fracture.In general,the mechanical property of alloy M5 is more stable than that of Zirlo alloy.
引文
[1]王峰.核级锆及锆合金研究状况及发展前景[J].兵器材料科学与工程,2012,35(1):107.WANG Feng.The Research Status and Development Prospects of Nuclear Grade Zirconium and Zirconium Alloy[J].Ordnance Material Science and Engineering,2012,35(1):107.(in Chinese)
[2]赵文金,周邦新.我国高性能锆合金的发展[J].原子能科学技术,2005,39(S1):2.ZHAO Wenjin,ZHOU Bangxin.High Performance Zirconium Alloys Development in China[J].Atomic Energy Science and Technology,2005,39(S1):2.(in Chinese)
[3]袁改焕,李恒羽,王德华.锆材在核电站的应用及前景[J].稀有金属快报,2007,26(1):14.YUAN Gaihuan,LI Hengyu,WANG Dehua.Application of Zirconium Material for Nuclear Power Station[J].Rare Metals Letters,2007,26(1):14.(in Chinese)
[4]张聪惠,刘大利,兰新哲,等.锆合金氧化膜腐蚀性能影响研究现状[J].材料热处理技术,2011,40(16):117.ZHANG Conghui,LIU Dali,LAN Xinzhe,et al.Study Progress about Effect of Oxide Film on Corrosion Resistance of Zircaloy[J].Material&Heat Treatment,2011,40(16):117.(in Chinese)
[5]陈宝山.我国压水堆核电燃料元件的发展[J].原子能科学技术,2003,37(S1):10.CHEN Baoshan.The Development of PWR Nuclear Fuel Element in China[J].Atomic Energy Science and Technology,2003,37(S1):10.
[6]马林生,王快社,岳强,等.核反应堆用锆合金性能分析[J].金属世界,2014(5):38.MA Linsheng,WANG Kuaishe,YUE Qiang.Performance Analysis of Zirconium Alloys for Neclear Reactor[J].Metal World,2014(5):38.
[7]袁改焕,卫新民.锆合金研究进展及我国核电站用锆材国产化的思考[J].钛工业进展,2011,28(6):18.YUAN Gaihuan,WEI Xinmin.The Research Progress of Zirconium Alloys and Think with Zirconium Localization of Nuclear Power Plant in China[J].Titanium Industry Progress,2011,28(6):18.(in Chinese)
[8]周军,李中奎.轻水反应堆(LWR)用包壳材料研究进展[J].中国材料进展,2014(9):554.ZHOU Jun,LI Zhongkui.Research Progress on Cladding Materials Used for Light Water Reactor[J].Materials China,2014(9):554.
[9]刘鹏,杜忠泽,马林生.核级锆及锆合金腐蚀性能研究现状[J].材料热处理技术,2011,40(22):22.LIU Peng,DU Zhongze,MA Linsheng.The Research Present Situation of Nuclear Gtade Zirconium and Zirconium Alloy in Corrosion Resistance[J].Material&Heat Treatment,2011,40(22):22.(in Chinese)
[10]周邦新,李强,黄强.水化学对锆合金耐腐蚀性能影响的研究[J].核动力工程,2000,2(5):441.ZHOU Bangxin,LI Qiang,HUANG Qiang.The Effect of Water Chemistry on the Corrosion Behavior of Zirconium Alloys[J].Nuclear Power Engineering,2000,2(5):441.(in Chinese)
[11]王正品,渠静雯,高巍.退火态及马氏体态Zr-Nb合金拉伸性能分析[J].西安工业大学学报,2013,33(4):319.WANG Zhengpin,QU Jingwen,GAO Wei.Study on Ring Tensile Properties of Martensitic and Annealed Zr-Nb Alloys[J].Journal of Xi’an Technological University,2013,33(4):319.(in Chinese)
[12]DURIEZ C,DUPONT T,SCHMET B,et al.Zircaloy-4and M5High Temperature Oxidation and Nitriding in Air[J].Journal of Nuclear Materials,2008,380(S1/3):30.
[13]张长义,宁广胜,佟振峰,等.M5锆合金包壳管轴向和环向拉伸性能测试[J].原子能科学技术,2005,39(S1):34.ZHANG Changyi,NING Guangsheng,TONG Zhenfeng,et al.Test on Tension Properties of M5 Alloy[J].Atomic Energy Science and Technology,2005,39(S1):34.
[14]CHATTERJEE S,SHAH P K,DUBEY J S.Ageing of Zirconium Alloy Components[J].Journal of Nuclear Materials,2008,383(S1/2):172.
[2]赵文金,周邦新.我国高性能锆合金的发展[J].原子能科学技术,2005,39(S1):2.ZHAO Wenjin,ZHOU Bangxin.High Performance Zirconium Alloys Development in China[J].Atomic Energy Science and Technology,2005,39(S1):2.(in Chinese)
[3]袁改焕,李恒羽,王德华.锆材在核电站的应用及前景[J].稀有金属快报,2007,26(1):14.YUAN Gaihuan,LI Hengyu,WANG Dehua.Application of Zirconium Material for Nuclear Power Station[J].Rare Metals Letters,2007,26(1):14.(in Chinese)
[4]张聪惠,刘大利,兰新哲,等.锆合金氧化膜腐蚀性能影响研究现状[J].材料热处理技术,2011,40(16):117.ZHANG Conghui,LIU Dali,LAN Xinzhe,et al.Study Progress about Effect of Oxide Film on Corrosion Resistance of Zircaloy[J].Material&Heat Treatment,2011,40(16):117.(in Chinese)
[5]陈宝山.我国压水堆核电燃料元件的发展[J].原子能科学技术,2003,37(S1):10.CHEN Baoshan.The Development of PWR Nuclear Fuel Element in China[J].Atomic Energy Science and Technology,2003,37(S1):10.
[6]马林生,王快社,岳强,等.核反应堆用锆合金性能分析[J].金属世界,2014(5):38.MA Linsheng,WANG Kuaishe,YUE Qiang.Performance Analysis of Zirconium Alloys for Neclear Reactor[J].Metal World,2014(5):38.
[7]袁改焕,卫新民.锆合金研究进展及我国核电站用锆材国产化的思考[J].钛工业进展,2011,28(6):18.YUAN Gaihuan,WEI Xinmin.The Research Progress of Zirconium Alloys and Think with Zirconium Localization of Nuclear Power Plant in China[J].Titanium Industry Progress,2011,28(6):18.(in Chinese)
[8]周军,李中奎.轻水反应堆(LWR)用包壳材料研究进展[J].中国材料进展,2014(9):554.ZHOU Jun,LI Zhongkui.Research Progress on Cladding Materials Used for Light Water Reactor[J].Materials China,2014(9):554.
[9]刘鹏,杜忠泽,马林生.核级锆及锆合金腐蚀性能研究现状[J].材料热处理技术,2011,40(22):22.LIU Peng,DU Zhongze,MA Linsheng.The Research Present Situation of Nuclear Gtade Zirconium and Zirconium Alloy in Corrosion Resistance[J].Material&Heat Treatment,2011,40(22):22.(in Chinese)
[10]周邦新,李强,黄强.水化学对锆合金耐腐蚀性能影响的研究[J].核动力工程,2000,2(5):441.ZHOU Bangxin,LI Qiang,HUANG Qiang.The Effect of Water Chemistry on the Corrosion Behavior of Zirconium Alloys[J].Nuclear Power Engineering,2000,2(5):441.(in Chinese)
[11]王正品,渠静雯,高巍.退火态及马氏体态Zr-Nb合金拉伸性能分析[J].西安工业大学学报,2013,33(4):319.WANG Zhengpin,QU Jingwen,GAO Wei.Study on Ring Tensile Properties of Martensitic and Annealed Zr-Nb Alloys[J].Journal of Xi’an Technological University,2013,33(4):319.(in Chinese)
[12]DURIEZ C,DUPONT T,SCHMET B,et al.Zircaloy-4and M5High Temperature Oxidation and Nitriding in Air[J].Journal of Nuclear Materials,2008,380(S1/3):30.
[13]张长义,宁广胜,佟振峰,等.M5锆合金包壳管轴向和环向拉伸性能测试[J].原子能科学技术,2005,39(S1):34.ZHANG Changyi,NING Guangsheng,TONG Zhenfeng,et al.Test on Tension Properties of M5 Alloy[J].Atomic Energy Science and Technology,2005,39(S1):34.
[14]CHATTERJEE S,SHAH P K,DUBEY J S.Ageing of Zirconium Alloy Components[J].Journal of Nuclear Materials,2008,383(S1/2):172.