国产316LN不锈钢在模拟AP1000一回路水环境中的疲劳行为

一回路水环境下的疲劳性能是核电站主管道设计寿命评估的重要参数。针对国产主管道材料316LN开展了模拟AP1000一回路水环境的低周疲劳试验,分析了疲劳行为和失效机理。研究结果表明:国产316LN峰值应力随应变幅的增大而增大,大应变幅试样在疲劳过程中先后发生了循环硬化、循环软化和失稳,而小应变幅试样在失稳前未发生明显的循环硬化和循环软化;在应变幅由0.2%逐渐增加至1.2%的过程中,疲劳周次从105逐渐降低至102;疲劳断口具有典型的疲劳断口特征,裂纹萌生于试样表面,以穿晶方式垂直于主应力方向扩展,裂纹扩展区具有典型的疲劳辉纹,辉纹上有菱形颗粒状腐蚀产物,环境辅助开裂机制倾向于氢致开裂。     

核电用316LN不锈钢的热机械疲劳性能研究

采用热机械疲劳试验方法研究316LN不锈钢的同相热机械疲劳行为,获得材料的疲劳数据。试验结果表明:316LN不锈钢的热机械疲劳行为是一个先强化后软化的过程;滞回曲线呈梭形,形状"饱满",具有良好的塑性变形能力,且随着温度范围增大,变形能力增强;在相同条件下,温度范围增大,材料的疲劳特征表现更为明显;在波动管运行条件下(温度≤320℃),应变对材料的疲劳寿命影响占主要作用;材料在120~320℃和120~230℃条件下的热机械疲劳寿命均大于350℃恒温低周疲劳寿命,说明采用传统的高温低周疲劳试验结果来评价波动管材料的热机械疲劳寿命过于保守。     

国产316LN不锈钢在AP1000一回路水环境的疲劳机理

正316LN不锈钢是第3代核电站AP1000的主管道材料,研究其在一回路水环境疲劳失效机理,对于主管道的安全和材料的改进具有重要的理论和现实意义。本文对316LN在模拟AP1000一回路水环境(321℃、15.5 MPa和0.1ppm)的疲劳失效样品进行了微观观察,分析了316LN在一回路水环境的疲劳机理。研究结果表明:316LN不锈钢在一回路水环境的疲劳裂纹主要源自于样品表面的驻留滑移带、夹杂物以及晶界(图1);裂纹扩展区断口(图2)具有脆性疲劳辉纹,辉纹上覆盖有菱形颗粒     

316不锈钢室温和350℃低周疲劳性能研究

采用MTS材料试验机研究作为反应堆结构材料的316奥氏体不锈钢母材在350℃和室温,以及焊缝在室温,±0.3%~1.5%应变幅的低周疲劳性能试验,并采用扫描电镜对试验后样品进行了断口分析。研究结果表明,316不锈钢疲劳性能较好,室温下疲劳寿命高出350℃同一应变幅的30%~50%以上,且母材的疲劳寿命显著高出焊缝同一应变幅的一倍以上。随应变幅的增加,材料疲劳寿命相应下降,峰值应力增加。室温下母材和焊缝均呈现出随循环周次增加、峰值随应力逐渐下降的规律。母材在高温下,随应变幅的增加,逐渐由循环硬化过渡到饱和行为。低周疲劳试验后,断口表面可观察到裂纹源和疲劳条带。随应变幅增加,疲劳条带间距增大,且同一应变幅下,焊缝的间距大于母材,高温的疲劳间距大于室温,与疲劳试验结果相吻合。     

国产F316Ti在模拟轻水堆一回路环境下的低周疲劳性能(英文)

轻水堆(LWR)环境对压力边界材料疲劳性能的影响,包括疲劳寿命及疲劳裂纹扩展速率对设备核安全是非常重要的。为了预测核材料疲劳寿命,改进核材料设计,对国产材料进行腐蚀环境下的疲劳性能研究以得到模拟环境下的疲劳数据是很有必要的。对国产F316Ti在模拟LWR一回路环境下的低周疲劳性能进行了研究,结果显示,高温水环境是影响奥氏体不锈钢低周疲劳性能的重要因素之一。对于同种材料,高温空气中的低周疲劳性能优于高温水中的低周疲劳性能;高温水中,国产F316Ti与日本奥氏体不锈钢具有同等的抗低周疲劳性能。腐蚀疲劳数据均处于ASME最佳拟合曲线和ASME设计疲劳曲线之间;F316Ti在模拟压水堆(PWR)和沸水堆(BWR)一回路环境中的低周疲劳性能,在高应变范围无明显差异;随着应变幅降低,渐见差异。模拟BWR环境中,数据处于较短寿命侧。由抚顺钢厂生产的F316Ti材料,钛均匀地分布于其中,且材料中Ni,Cr,Mo含量均处于合金化学成分上限。因此,它具有较优越的抗高温水的腐蚀疲劳性能。     

SCWR候选不锈钢的高周疲劳行为研究

采用旋转弯曲的加载方式对奥氏体不锈钢347、316Ti、310进行疲劳试验。试验环境为室温下空气中和550℃空气中。对疲劳断口进行扫描电镜(SEM)分析,根据试验数据绘制材料的应力-循环(S-N)曲线。结果表明,3种不锈钢疲劳极限大小顺序为347<316Ti<310,与静强度顺序一致;高温会加速试样的氧化,降低材料的疲劳寿命,347不锈钢的下降趋势最大,对温度最敏感;疲劳极限试验与经验公式计算值的比较表明,3种不锈钢具有较好的抗高周疲劳性能;疲劳过程为裂纹源产生、扩展和断裂,疲劳条带宽度在1μm左右的量级,最后断裂区具有韧窝特征,347不锈钢的韧窝中分布着数量较多的大小孔洞。     

国产A508-3钢在模拟AP1000一回路水环境下的疲劳性能研究

反应堆压力容器（RPV）钢在一回路水环境下的疲劳性能是评价其设计寿命的重要参数。本文针对国产A508-3钢开展了模拟AP1000一回路水环境的低周疲劳性能试验研究，获得了321 ℃、155 MPa及01 ppm溶解氧水环境下的疲劳行为数据和断裂机理。研究结果表明，国产A508 3钢峰值应力随应变幅的增大而逐渐增大，疲劳试验过程中试样表现出循环硬化、循环软化和饱和3个阶段；在应变幅由02%逐渐增加至06%的过程中，疲劳周次从105逐渐降低至102；疲劳断口具有疲劳和腐蚀特征，属于典型的腐蚀疲劳断裂。     

304不锈钢的单轴棘轮变形与失效行为研究

对304不锈钢在室温下进行了单轴应变控制下的应变棘轮变形与失效以及低周疲劳试验研究,系统地揭示了材料在循环过程中的材料变形与失效行为。研究表明：材料的应变棘轮变形与失效既不同于单轴拉伸,也不同于相同应变幅值下的对称应变循环加载时的变形与失效,而是强烈地依赖于应变幅值与每一循环周次在最大拉应变处的应变增加量。观察到了一些有意义的结果,     

N18合金薄壁管高温应变循环与疲劳行为研究

应用新型自研夹具对N18合金薄壁短管进行400℃下的单轴拉伸和等幅低周应变疲劳试验。试验结果表明：N18短管高温循环应力应变滞回线有良好对称性；等幅循环下短管试样在较低应变幅下表现出循环硬化特性，而在较高应变幅下表现出循环软化；在多级应变循环加载下短管试样应力幅在循环中均保持稳定，循环本构关系不受多级应变循环工况差异的影响；材料循环特性不符合Manson律。获得了用于N18合金在400℃高温下的几个寿命估算式。     

国产F316Ti在模拟轻水堆一回路环境下的低周疲劳性能

轻水堆(LWR)环境对压力边界材料疲劳性能的影响,包括疲劳寿命及疲劳裂纹扩展速率对设备核安全是非常重要的。为了预测核材料疲劳寿命,改进核材料设计,对国产材料进行腐蚀环境下的疲劳性能研究以得到模拟环境下的疲劳数据是很有必要的。对国产F316Ti在模拟LWR一回路环境下的低周疲劳性能进行了研究,结果显示,高温水环境是影响奥氏体不锈钢低周疲劳性能的重要因素之一。对于同种材料,高温空气中的低周疲劳     

Development of thermal fatigue testing apparatus with BWR water environment and thermal fatigue strength of austenitic stainless steels

A thermal fatigue testing apparatus was developed in order to clarify the fatigue behavior in BWR environment. Pressurized high and low temperature pure water were alternately supplied into an autoclave with a small cylindrical specimen. Then a fatigue specimen was subjected to homogeneous thermal stress through the wall thickness. Fatigue crack initiation behavior was observed with the replication method and compared with the mechanical fatigue strength performed in air and high temperature water. The thermal fatigue strength of type 304 and 316 nuclear grade (316NG) stainless steels agreed closely with the mechanical fatigue strength, when transforming the nominal stress amplitude to the fictitious stress amplitude by using the mean value of strain amplitudes for room temperature and 288°C.     

Fatigue Life and Crack Growth Mechanisms of the Type 316LN Austenitic Stainless Steel in 310°C Deoxygenated Water

The low cycle fatigue tests of the type 316LN stainless steel were conducted to investigate the cracking mechanisms in high-temperature water. The fatigue lives of the specimens tested in 310°C deoxygenated water were considerably shorter than those tested in air. For the specimens tested in 310°C deoxygenated water, the evidences for the metal dissolution such as the stream downed feature, the blunt crack shape, and the wider crack opening were observed but rather weakly. In the same specimens, the evidences for the hydrogen-induced cracking such as the coalescence of microvoids and the decrease of the dislocation spacing at the crack tip were observed rather clearly. Therefore, it is thought that the hydrogen-induced cracking is mainly responsible for the reduction in the fatigue life of the type 316LN stainless steel in 310°C deoxygenated water while the effect of metal dissolution is less significant. The hydrogen-induced cracking is more pronounced in the slower strain rates. This behavior is in accordance with the larger reduction in the fatigue life at the slower strain rates. Furthermore, the fatigue life and the dislocation spacing show the minimum value in the strain rate range from 0.008 to 0.04%/s, which indicates the existence of the critical strain rate.     

高温水中氯离子对316L不锈钢应力腐蚀裂纹扩展速率的影响

本文采用直流电压降(DCPD)方法,使用恒K(K=27.5 MPa·m1/2)加载方式,在核电厂高温高压水环境中研究了氯离子对316L不锈钢的应力腐蚀裂纹扩展速率的影响。实验结果表明:在高温除氧水中,氯离子会加快316L不锈钢的应力腐蚀裂纹扩展速率,且当水中存在溶解氧时,氯离子对应力腐蚀裂纹扩展速率的影响更明显。     

Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions

The presented paper summarizes the results of general corrosion and stress corrosion cracking (SCC) susceptibility tests in supercritical water (SCW), studied for austenitic stainless steel 316L, with the aim to identify maximum SCW temperature usability and specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralized SCW solution with controlled oxygen content. The general corrosion tests clearly revealed the applicability of austenitic stainless steel in SCW to be limited to 550 °C as maximum temperature as oxidation rates of austenitic stainless steels 316L increase dramatically above 550 °C. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 550 °C SCW. Besides the strain rate (resp. crosshead speed), the oxygen content was varied in the series of tests. The obtained results showed that even at the lowest strain rate, a serious increase of SCC susceptibility, as typically characterized by IGSCC crack growth, was not observed. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. Based on fractographic findings a phenomenological map describing the SCC regime of SSRT test parameters could be proposed for AISI 316L.     

应变幅对国产锻造奥氏体不锈钢环境疲劳寿命影响的试验研究

为验证模拟压水堆核电站冷却剂服役环境对国产锻造主管道用奥氏体不锈钢疲劳寿命的影响,采用高温高压循环水疲劳测试系统对从产品锻件取样加工后的标准试样进行了低周疲劳试验,分析了试验数据与美国机械工程师学会(American Society of Mechanical Engineers,ASME)规范平均/设计疲劳曲线的关系,获得了应变幅对奥氏体不锈钢环境疲劳寿命的影响规律,并初步评价了ASME规范设计疲劳曲线和环境疲劳修正系数的适合性。     

焊接残余应力对316LN不锈钢应力腐蚀裂纹扩展速率的影响

核电站不锈钢管道焊接过程中引入的残余应力对焊接接头的应力腐蚀开裂性能有较大影响。本文针对一AP1000主管道316LN不锈钢焊接模拟件进行残余应力分析和应力腐蚀裂纹扩展速率测量,得到了焊后原始状态和去应力热处理状态的焊接热影响区材料在高温高压水中的应力腐蚀裂纹扩展速率。实验结果表明,焊接残余应力明显提高了热影响区的应力腐蚀裂纹扩展速率,且在含氢的压水堆一回路正常水化学下焊接残余应力的影响更加显著。     

Effect of ratcheting deformation on fatigue and creep-fatigue life of 316FR stainless steel

Components of fast breeder reactor (FBR) plants will be subjected to large thermal load, and progressive deformation with loading cycles (ratcheting) and creep-fatigue damage should be considered in their design. To clarify the effect of ratcheting on fatigue and creep-fatigue life, a series of fatigue and creep-fatigue tests coupled with strain progress were carried out for 316FR stainless steel. It was found that tensile ratcheting decreases the failure life to a large extent at small strain range, while compressive ratcheting does not decrease the failure life. Measurement of striation intervals on fracture surface showed small influence of strain increment on the crack propagation rate, suggesting that the main cause of the life reduction is the decrease in the crack initiation life. It was also found that failure life in various conditions is correlated well with a product of strain range and tensile peak stress.     

Effects of loading factors on environmental fatigue behavior of low-alloy pressure vessel steels in simulated BWR water

Low cycle fatigue resistance of low-alloy pressure vessel steels was investigated in simulated boiling water reactor (BWR) water. Much attention was paid to the effects of loading factors on fatigue life and environmentally assisted cracking (EAC) behavior, in which strain rate, strain waveform and strain amplitude were taken into account. The fatigue resistance and EAC behavior of the steels in simulated BWR water were found to be closely dependent on the strain rate, strain waveform and strain amplitude applied. The above fatigue behavior may be attributed to loading-factor-induced change in dominant EAC processes in high temperature water environments. Related EAC mechanisms are also discussed.     

The effect of a single tensile overload on stress corrosion cracking growth of stainless steel in a light water reactor environment

It has been found that a single tensile overload applied during constant load amplitude might cause crack growth rate retardation in various crack propagating experiments which include fatigue test and stress corrosion cracking (SCC) test. To understand the affecting mechanism of a single tensile overload on SCC growth rate of stainless steel or nickel base alloy in light water reactor environment, based on elastic-plastic finite element method (EPFEM), the residual plastic strain in both tips of stationary and growing crack of contoured double cantilever beam (CDCB) specimen was simulated and analyzed in this study. The results of this investigation demonstrate that a residual plastic strain in the region immediately ahead of the crack tips will be produced when a single tensile overload is applied, and the residual plastic strain will decrease the plastic strain rate level in the growing crack tip, which will causes crack growth rate retardation in the tip of SCC.