CF-8M铸造不锈钢主管道的热老化脆化行为研究

为了评价和预测铸造奥氏体不锈钢(CASS)材料服役期限内的热老化脆化程度,通过对美国的阿贡实验室(ANL)预测模型的研究和分析,以及在400℃下对核级CF-8M主管道材料实施了10 000 h的加速热老化试验,研究了CF-8M材料在不同热老化时间下拉伸性能、冲击性能和微观组织的变化规律,以冲击能作为表征热老化脆化程度的参数,获得了CF-8M材料的热老化脆化预测关系式,并与ANL模型的预测结果进行了对比和分析。结果表明,在加速热老化试验周期内随着热老化时间的增加,CF-8M材料的室温和高温(350℃)0.2%塑性延伸强度变化缓慢,抗拉强度缓慢增加;室温冲击能迅速下降,8 000 h以后冲击能下降趋势接近饱和状态;ANL模型对试验对象在加速老化试验周期内的冲击能预测结果不保守。CF-8M材料加速热老化10 000 h即等效服役30.49 a,其热老化脆化程度接近于热老化饱和状态。     

基于磁性多参数的铸造奥氏体不锈钢热老化状态评估研究

研究了应用磁性无损检测方法评估核电站主管道用铸造奥氏体不锈钢 (CASS)的热老化状态,结果表明在400 ℃下热老化不同时间,CASS的冲击功随着热老化时间延长逐渐降低,受材料组织不均应等因素影响,单一磁性参数与CASS的热老化时间之间无对应的单调规律;以磁性多参数测试为基础,结合主成分分析和非线性多元回归对测得的磁性...     

核电用铸造奥氏体不锈钢的热老化性能研究

压水堆核电厂主管道用国产铸造奥氏体不锈钢(CASS)长期服役会面临着热老化问题。在400℃高温下开展CASS材料的加速热老化实验,采用夏比冲击试验获得了材料室温吸收能量随热老化时间延长的下降规律,采用扫描电镜观察到不同热老化时间冲击断口形貌的变化趋势。实验结果表明:经过15 000h的加速热老化实验,CASS材料的热老化程度逐渐达到饱和状态,吸收能量虽大幅下降,但仍能满足设计规范对CASS材料在未老化状态时的考核要求。     

Z3CN20.09M奥氏体不锈钢热老化冲击性能试验研究

采用GB/T19748-2005钢材夏比V型缺口摆锤冲击试验仪器化试验方法,对压水堆核电厂用离心铸造Z3CN20.09M奥氏体不锈钢主管道样品进行了实验室热老化的冲击性能研究。冲击试验数据的统计分析表明,热老化对Fiu/Fm比值不产生影响,而对冲击载荷有显著影响,对冲击能量的影响则更为显著。透射电子显微分析表明,热老化导致铁素体中出现沉淀物,并引发了奥氏体中位错组态的改变。与热老化时间lg t之间也满足线性关系。     

沉淀硬化马氏体不锈钢热老化脆化的热电势检测研究

研究了17-4PH沉淀硬化马氏体不锈钢在400℃下不同时效时间力学性能和热电势变化规律,提出了热电势和冲击韧性的经验公式,并通过在核电厂服役13 a的主蒸汽隔离阀阀杆进行验证。验证结果显示,随着时效时间的延长,材料的冲击韧性下降,屈服强度、抗拉强度和硬度升高,断面收缩率和断裂伸长率下降。材料的热电势变化与冲击韧性呈现指数相关性,材料的屈服强度、抗拉强度、硬度和热电势呈现出较好的线性关系。通过热电势检测评估的冲击韧性和实测值显示出较好的符合性。      

316LN奥氏体不锈钢焊缝低温热老化行为研究

在压水堆核电厂中,主管道奥氏体不锈钢焊缝长期在其热老化敏感温度(280~325℃)下运行,为了研究主管道奥氏体不锈钢焊缝在核电厂运行温度下的热老化性能,开展了铁素体含量为10.7%的316LN不锈钢主管道焊缝在325、365、400℃下的低温热老化行为研究。结果表明:经6000 h热老化后,焊缝中铁素体相和奥氏体相中的主要元素含量没有发生明显变化,焊缝显微硬度快速增加但奥氏体相显微硬度没有发生变化,焊缝冲击功显著下降、拉伸性能变化较小。     

轻水堆核电站奥氏体不锈钢铸件的热老化及其老化管理

在轻水堆核电站中,奥氏体不锈钢铸件(CASS)在运行温度下长期工作将面临热老化问题。本文对热老化的机理进行了描述,并针对核电站的热老化问题,给出了老化管理工作的主要步骤,即部件敏感性甄别、老化状态评估和ISI大纲更新;结合CASS热老化的老化管理实践,对我国核电站的热老化管理工作提出了建议。     

热老化对20Cr25NiNb不锈钢冲击韧性的影响研究

针对长时间高温下合金力学性能退化问题,开展超临界气冷堆候选包壳材料的热老化研究。对改进型气冷堆用原型20Cr25NiNb不锈钢和添加不同元素的改进型合金,开展650℃下3000 h热老化试验。组织和性能结果表明,所有合金的冲击吸收能量(KV₂)均随热老化进行而下降。这种塑性降低与高温下第二相演化密切相关。沿晶界先后析出M₂₃C₆和G相导致原型合金冲击韧性先下降再缓慢上升。添加W和Mo元素后,沿晶界析出Laves和σ相,引起KV₂下降更快;B元素可细化晶界σ相,使得冲击韧性下降幅度小于不含B元素。加入Al元素后,合金基体中析出大量Laves和NiAl相,同时晶界σ相快速粗化,导致材料脆化严重。     

核电站一回路奥氏体不锈钢铸件热老化的微观力学性能变化

在核电站一回路中,奥氏体不锈钢铸件(CASS)在运行温度下长期工作除了受到辐照损伤外.还将面临热老化问题.本文对不同时间热老化后的CASS样品进行微观力学性能分析,并结合微观组织中铁素体的含量及分布对CASS的性能变化进行了分析,结果表明微观力学性能的变化和常规的宏观力学性能变化的规律不尽相同.丰富了CASS经老化后的力学性能数据,为CASS的安全运行提供参考.     

核电厂主管道90°弯头热老化程度的评估

根据IAEA和国际上相关文献提出的铸造奥氏体-铁素体不锈钢热老化评价方法和评估流程,以秦山核电厂二期扩建工程为例,对静态铸造奥氏体-铁素体不锈钢主管道90°弯头在长时间热老化后,以及在325℃服役10 a后的冲击韧性和断裂韧性J-R曲线进行评估。评估结果表明:秦山核电厂二期扩建工程主管道静态铸造奥氏体-铁素体不锈钢主管道90°弯头在325℃服役10 a后的冲击韧性和断裂韧性有明显的下降,但冲击韧性仍在安全裕量以内,长时间热老化后的冲击韧性已低于设计要求。     

Thermal aging effect on Z3CN20.09M Cast Duplex Stainless Steel

Instrumented impact tests were performed on Z3CN20.09M Cast Duplex Stainless Steel (CDSS) aged at 400 °C for up to 3000 h. The material was cut from the primary pipe of a PWR to investigate the accelerated thermal aging embrittlement effect on the dynamic fracture properties. The load–deflection curves from the instrumented impact tests described both the loading and the fracture stages, from which the dynamic ultimate loading level and the dynamic ultimate strengths were calculated as a function of increased aging time. The increase in the dynamic ultimate strengths with aging time is much higher than those for the static ultimate strengths. With the increment of the aging time, both the crack initiation energy and the crack propagation energy decrease, which are separated form the total impact energy taking the maximum loading level as the crack initiation point. It is shown that the reduction tendency of the fracture propagation energy is the predominant reason for the reduction of the impact energy. With the increment of the aging time, the unloading slopes reflecting the crack propagation speed increase and the fracture surface patterns change from ductile fracture with shallow dimples to cleavage brittle fracture caused by dislocations piling up in the ferrite matrix. Mechanical properties of inner wall with fine equi-axed grains are much better than those of the outer wall with coarse columnar structures.     

Investigation of thermal aging damage mechanism of the Cast Duplex Stainless Steel

Besides the macro-mechanical properties for thermal aging effect published in “Thermal aging effect on Z3CN20.09M Cast Duplex Stainless Steel” (Nuclear Engineering and Design 239(2009) 2217-2223), the thermal aging damage mechanism is investigated in this paper through nano-indentation tests and micro-structures evolution examination. Numerical simulations were carried out with GTN continuum damage model to investigate the different crack propagation process for aging. The nano-indentation hardness values increase with aging time for both phases while the hardness values of the ferrite phase are much higher and increase much more. The nano-indentation energy indicating the toughness decreases for both phases with aging time. TEM results show that the Cr-enriched α′ phase precipitates in the ferrite phase which is considered as the critical reason making the dislocation slip difficult and causing the increase of the strength and reduction of the toughness. The crack initiates from the ferrite phase instead of the austenite phase from the SEM observation and FEA simulation results, which reflects the change of the fracture mechanism for thermal aging.     

Small-angle Scattering Study of Structure of 17-4PH Stainless Steel Valve Stem Due to Thermal Aging Effect

<正>The thermal aging effect of 17-4PH stainless steel stem used in a nuclear power plant(NPP) in China after long-term service at about 300℃have been studied by small-angle neutron scattering (SANS) technique,atomic probe tomography(APT),electron microscopy (EM),impact ener gy test and also theoretical simulation.     

Effect of Element Segregation on Thermal Aging Behavior of 17-4PH Martensitic Stainless Steel for Nuclear Power Plant

正For thermal aging behavior of 17-4PH martensitic stainless steel,the former results show that the mechanical properties will be significant degeneration under long-time service at high temperature(~300℃).It exhibits a decline of DBTT and USE,with hardness increased and plasticity     

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。     

316不锈钢长时总体一次薄膜应力强度许用值预测方法研究

材料的总体一次薄膜应力强度许用值(S_mt)是高温反应堆设备结构设计力学分析的重要判定依据,但美国机械工程师协会(ASME)的规范和法国《快堆核岛机械设备设计和建造规范》(RCC-MR)给出的最长30万小时的S_mt不能满足长寿期反应堆的设计要求。本文基于ASME规范给出的30万小时许用应力、预计最小断裂应力及断裂应力系数等材料蠕变性能数据,采用Larson-Miller外推模型成功获得了50万小时长寿期的316不锈钢母材和焊缝的长时蠕变性能,可满足长寿期反应堆的设计要求。