反应堆压力容器用Ni-Cr-Mo-V钢焊缝热老化脆化行为研究
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摘要
对反应堆压力容器用Ni-Cr-Mo-V钢焊缝温度监督样品的热老化脆化行为进行了研究。焊缝属于压力容器的薄弱环节,服役时间最高达120 430h(服役温度归一化到300℃)。3批次的焊缝监督样品冲击实验表明,焊缝材料在热老化过程中发生了脆化。通过研究发现,金相组织和显微维氏硬度在热老化期间未发生明显的变化,表明在热老化过程中不存在硬化脆化机制。断口分析及扫描俄歇纳米探针研究表明,晶界发生了P的偏析,弱化了晶界结合力,因此,反应堆压力容器用Ni-Cr-Mo-V钢焊缝在热老化过程中发生了由杂质元素P偏析引起的非硬化脆化。
The study of thermal ageing embrittlement of temperature sets of reactor pressure vessel surveillance Ni-Cr-Mo-V steel weld metal was conducted.Weld metal is the weakness of the reactor pressure vessel and its service time gets to 120 430 h(the operating temperature is normalized to 300℃).The impact test of the 3 batches of weld surveillance sample indicates that the weld metal is embrittled during thermal ageing.Through the study,significant change is not found in the microstructure and microVickers hardness during thermal ageing,indicating that there is no hardening embrittlement mechanism in the thermal ageing process.The study of impact fracture and Auger electron spectroscopy indicates that the impurity element P segregates to thegrain boundaries and lowers their cohesion during the long-term thermal ageing.Therefore,the reactor pressure vessel Ni-Cr-Mo-V steel weld metal occurs non-hardening embrittlement which is probably caused by the intergranular segregation of impurity element P during the long-term thermal ageing.
The study of thermal ageing embrittlement of temperature sets of reactor pressure vessel surveillance Ni-Cr-Mo-V steel weld metal was conducted.Weld metal is the weakness of the reactor pressure vessel and its service time gets to 120 430 h(the operating temperature is normalized to 300℃).The impact test of the 3 batches of weld surveillance sample indicates that the weld metal is embrittled during thermal ageing.Through the study,significant change is not found in the microstructure and microVickers hardness during thermal ageing,indicating that there is no hardening embrittlement mechanism in the thermal ageing process.The study of impact fracture and Auger electron spectroscopy indicates that the impurity element P segregates to thegrain boundaries and lowers their cohesion during the long-term thermal ageing.Therefore,the reactor pressure vessel Ni-Cr-Mo-V steel weld metal occurs non-hardening embrittlement which is probably caused by the intergranular segregation of impurity element P during the long-term thermal ageing.
引文
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[2]MARGOLIN B Z,NIKOLAYEV V A,YURCHENKO E V,et al.Analysis of embrittlement of WWER-1000 RPV materials[J].International Journal of Pressure Vessels&Piping,2012,89:178-186.
[3]ASTM E185 standard practice for design of surveillance programs for light-water moderated nuclear power reactor vessels[S].USA:ASTM International,2002.
[4]GORITSKII V M,SHNEIDEROV G R,SHUR A D,et al.Structural mechanism of the development of thermal embrittlement in steels with a temper-bainite structure[J].Metal Science and Heat Treatment,1992,34(1):3-9.
[5]SHTROMBAKH Y I,GUROVICH B A,KULESHOVA E A,et al.Thermal ageing mechanisms of VVER-1000reactor pressure vessel steels[J].Journal of Nuclear Materials,2014,452(1):348-358.
[6]KULESHOVA E A,GUROVICH B A,LAVRUKHINA Z V,et al.Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation[J].Journal of Nuclear Materials,2016,477:110-122.
[7]UTEVSKIY L M.Temper brittleness of steel[J].Metal Science and Heat Treatment of Metals,1961,3(5-6):263-267.
[8]佟振峰,崔贞北,赵继松,等.VVER-1000型反应堆压力容器热老化分析评估[J].原子能科学技术,2015,49(5):903-908.TONG Zhenfeng,CUI Zhenbei,ZHAO Jisong,et al.Thermal ageing assessment of RPV in VVER-1000reactor[J].Atomic Energy Science and Technology,2015,49(5):903-908(in Chinese).
[9]Central Institute of Atomic Energy Information Science.PNAE G-7-002-86 Strength calculations norms for nuclear power plant equipment and piping[S].Moscow:Energoatomizdat,1989.
[10]National Standards Board under Council of Ministers of the USSR.GOST 9454-78 Test method for testing the impact strength at low,room and high temperature[S].Moscow:Izd standartov,1985.
[11]迟之东.对接焊缝金属和熔敷金属冲击功差异原因探讨[D].哈尔滨:哈尔滨工程大学,2008.
[12]薛屺,刘德臣,尹长华,等.焊缝中M-A组元对X70管线钢焊缝组织及冲击性能的影响[J].金属热处理,2013,38(8):32-37.XUE Qi,LIU Dechen,YIN Changhua,et al.Influence of M-A constituent on microstructure and impact property of weld metal of X70pipeline steel[J].Heat Treatment of Metals,2013,38(8):32-37(in Chinese).
[13]GUROVICH B,KULESHOVA E,ZABUSOV O,et al.Influence of structural parameters on the tendency of VVER-1000reactor pressure vessel steel to temper embrittlement[J].Journal of Nuclear Materials,2013,435(1):25-31.
[14]WU J,SONG S H,WENG L Q,et al.An Auger electron spectroscopy study of phosphorus and molybdenum grain boundary segregation in a2.25Cr1Mo steel[J].Materials Characterization,2008,59(3):261-265.
[15]徐庭栋.非平衡晶界偏聚动力学和晶间脆性断裂[M].北京:科学出版社,2006.
[16]HONDROS E D,SEAH M P.The theory of grain boundary segregation in terms of surface adsorption analogues[J].Metallurgical and Materials Transactions A,1977,8(9):1 363-1 371.
[17]SEAH M P.Adsorption-induced interface decohesion[J].Acta Metallurgica,1980,28(7):955-962.