高温He离子辐照GH3535合金的损伤效应

采用500 keV的He离子在750 ℃下对GH3535合金样品进行辐照,然后利用掠入射X射线衍射（GIXRD）、透射电子显微镜（TEM）和纳米压痕仪分别对样品的氦泡和位错环辐照缺陷的演化及纳米硬度的变化进行了研究。结果表明,GH3535合金晶格辐照后发生了轻微畸变;离子辐照在样品中形成了大量尺寸为2～5 nm的氦泡和位错环。辐照产生的氦泡和位错环等缺陷在基体中钉扎位错,从而使材料产生了辐照硬化现象,样品硬度随辐照剂量的增加而增大。当辐照剂量达2×1016 cm-2时,辐照样品发生了明显的硬化饱和现象,利用Nix Gao模型计算得此时的硬化程度为64%。     

高温He离子辐照GH3535合金的损伤效应

采用500 keV的He离子在750℃下对GH3535合金样品进行辐照,然后利用掠入射X射线衍射(GIXRD)、透射电子显微镜(TEM)和纳米压痕仪分别对样品的氦泡和位错环辐照缺陷的演化及纳米硬度的变化进行了研究。结果表明,GH3535合金晶格辐照后发生了轻微畸变;离子辐照在样品中形成了大量尺寸为2～5 nm的氦泡和位错环。辐照产生的氦泡和位错环等缺陷在基体中钉扎位错,从而使材料产生了辐照硬化现象,样品硬度随辐照剂量的增加而增大。当辐照剂量达2×10~(16) cm~(-2)时,辐照样品发生了明显的硬化饱和现象,利用Nix-Gao模型计算得此时的硬化程度为64%。     

辐照硬化位错动力学模拟的研究进展

辐照硬化是金属材料的辐照效应之一,开展辐照硬化机理研究有助于设计可靠的反应堆结构材料。辐照产生的缺陷会对位错运动造成阻碍,被认为是辐照硬化的主要原因。近年来快速发展的位错动力学模拟方法为材料的微观组织变化和宏观力学性能之间建立起了桥梁,被广泛用于辐照硬化机理研究。对于一些辐照缺陷如位错环和层错四面体,位错动力学软件已能模拟它们对位错网络演化以及宏观力学响应的影响,使辐照硬化的定量预测成为可能。本文从位错动力学模型、不同类型辐照缺陷硬化效应的位错动力学模拟以及辐照硬化理论模型发展三个方面,综述了辐照硬化位错动力学模拟的研究进展,并展望该研究领域的主要科学问题。     

FeNiCr等比多主元固溶体合金中弗兰克位错环与位错相互作用研究

<正>多主元固溶体合金凭借其优秀的高温力学性能与抗辐照性能,在第4代反应堆研究中得到广泛关注。很多文献报道了辐照后多主元固溶体合金的高温力学性能和辐照损伤研究工作。由于辐照会引入位错环、空洞和析出物等辐照缺陷,这些缺陷与位错相互作用会进一步影响多主元固溶体合金的辐照硬化程度,但相关机理目前尚不清楚。本文采用分子动力学方法对FeNiCr等比多主元     

ODS-Eurofer钢微观结构及辐照硬化研究

研究了ODS-Eurofer钢的微观结构及辐照硬化现象。首先用透射电子显微镜(TEM)观察了ODS-Eurofer钢的初始微观组织结构,发现基体中不仅存在几nm至几十nm的氧化物弥散颗粒,还存在具有壳 核结构的大尺寸(直径大于100 nm)颗粒,并观察到纳米颗粒对位错线的钉扎作用。随后用能量为5 MeV的Fe²⁺离子在300 ℃和500 ℃下辐照样品至25 dpa以模拟中子辐照,并用纳米压痕仪和TEM测试表征了辐照所致力学性能和微观结构的变化。结果表明,两种温度下辐照均引起硬度上升,500 ℃时由于辐照产生的点缺陷发生复合,导致硬化效应弱于300 ℃。用TEM观测辐照水平为25 dpa的损伤层发现有少量纳米尺寸位错环,这些位错环是辐照硬化的主要原因。ODS-Eurofer钢初始微观结构对辐照硬化有重要影响,其中晶界、纳米颗粒与基体界面、位错线等能捕获辐照过程中产生的点缺陷,从而抑制辐照位错环的生长。     

ODS-Eurofer钢微观结构及辐照硬化研究

研究了ODS-Eurofer钢的微观结构及辐照硬化现象。首先用透射电子显微镜(TEM)观察了ODS-Eurofer钢的初始微观组织结构,发现基体中不仅存在几nm至几十nm的氧化物弥散颗粒,还存在具有壳-核结构的大尺寸(直径大于100 nm)颗粒,并观察到纳米颗粒对位错线的钉扎作用。随后用能量为5 MeV的Fe~(2+)离子在300℃和500℃下辐照样品至25 dpa以模拟中子辐照,并用纳米压痕仪和TEM测试表征了辐照所致力学性能和微观结构的变化。结果表明,两种温度下辐照均引起硬度上升,500℃时由于辐照产生的点缺陷发生复合,导致硬化效应弱于300℃。用TEM观测辐照水平为25 dpa的损伤层发现有少量纳米尺寸位错环,这些位错环是辐照硬化的主要原因。ODS-Eurofer钢初始微观结构对辐照硬化有重要影响,其中晶界、纳米颗粒与基体界面、位错线等能捕获辐照过程中产生的点缺陷,从而抑制辐照位错环的生长。     

辐照硬化位错动力学模拟的研究进展

辐照硬化是金属材料的辐照效应之一,开展辐照硬化机理研究有助于设计可靠的反应堆结构材料.辐照产生的缺陷会对位错运动造成阻碍,被认为是辐照硬化的主要原因.近年来快速发展的位错动力学模拟方法为材料的微观组织变化和宏观力学性能之间建立起了桥梁,被广泛用于辐照硬化机理研究.对于一些辐照缺陷如位错环和层错四面体,位错动力学软件已能...     

FeCr合金中Cr含量对微观结构影响的原子尺度模拟研究

低温辐照脆化是影响铁素体/马氏体（F/M）钢服役的主要问题之一。F/M钢低温辐照脆化的主要机理是辐照产生的纳米缺陷（如位错环、α′相(富Cr团簇)等）阻碍位错运动。本文利用分子动力学方法和迈氏蒙特卡罗方法对F/M钢模型材料--FeCr合金（Fe7%Cr、Fe9%Cr、Fe14%Cr）中Cr元素析出成团簇及在位错环上偏析的机理进行研究,并分析Cr团簇析出与合金成分的关系以及位错环尺寸、位错环类型和合金中Cr含量对位错环上Cr偏析量的影响。模拟结果表明：热力学模拟后,高Cr含量（>9%)的FeCr合金中会析出Cr团簇,且基体内Cr含量越高,析出的Cr团簇尺寸越大;在所研究的3种FeCr合金中,受位错环张应力场作用,合金元素Cr均会在位错环的外围偏析,且FeCr合金中Cr含量越高,Cr在位错环上偏析量越高。低Cr的FeCr合金中Cr对其辐照硬化的影响需考虑位错环上Cr偏析的影响,高Cr的FeCr合金中Cr元素对其辐照硬化的影响需综合考虑Cr团簇及位错环上Cr偏析。     

中子辐照对Cr涂层锆合金力学性能的影响

为研究中子辐照对Cr涂层锆合金力学性能的影响,获得Cr涂层锆合金的辐照性能数据,本文针对多弧离子镀技术制备的Cr涂层锆合金开展了中子辐照考验,通过拉伸试验过程实时观测试样力学行为变化并对试验后断口微观形貌进行分析,研究了辐照后Cr涂层锆合金的力学性能以及涂层与基体的结合能力。结果表明：中子辐照导致Cr涂层锆合金的抗拉强度和屈服强度升高,断后伸长率下降,表现出与商用锆合金相似的辐照强化效应。同时Cr涂层与无涂层锆合金相比,其屈服强度和抗拉强度升高但塑性变形能力降低。另一方面,Cr涂层在拉伸变形量较大时产生环向裂纹,但未从基体表面剥落,中子辐照未对涂层结合强度产生明显的影响,受力过程中涂层仍保持了完整性。     

电子辐照条件下高纯铁中位错环演化的多尺度模拟

辐照诱导材料微观结构演化导致的材料力学性能降级或尺寸不稳定性是限制反应堆安全与经济性的关键因素之一。本文基于速率理论建立了辐照诱导材料微观结构演化的物理模型,并开发了模拟程序Radieff。采用分子动力学计算了高纯铁中缺陷的形成能、结合能、迁移能以及间隙原子位错环的构型,在此基础上模拟了电子辐照诱导高纯铁内位错环的演化过程,并与实验结果进行了对比。基于分子动力学的计算结果表明,当间隙原子团簇包含3个间隙原子时,团簇的排列方式为〈110〉构型,间隙原子团簇包含4个以上间隙原子时,团簇排列方式变为〈111〉构型。此外基于Radieff研究了400~600K温度范围内,损伤速率为1.5×10-4 dpa/s电子辐照条件下,位错密度对位错环演化的影响,位错密度对位错环数密度及其平均尺寸的影响取决于位错以及间隙原子团簇对间隙原子的阱强度;在464K和550K温度下辐照,位错环数密度及其平均尺寸分别在位错密度增加到1011 cm-2和1010 cm-2后急剧减小,这是由于此时位错对间隙原子的阱强度会大于间隙原子团簇对间隙原子的阱强度。     

C276合金的抗辐照性能研究

C276合金为包壳部件的候选材料之一，本文拟对其抗辐照性能进行研究。对C276合金进行质子及多束粒子辐照，利用纳米硬度仪、透射电镜、拉曼成像仪等研究了C276合金在辐照前后的试样。结果表明：在质子及多束粒子辐照下，辐照损伤区域发生C偏析和位错环硬化；在H或He单束辐照条件下，在35.0 μm或3.5 μm深度处，拉曼光谱中的碳峰相对强度较大且碳峰红移，引起此处的纳米硬度较其他深度处的高；试验得到的损伤峰对应的深度与模拟计算得到的吻合。可推知，C276合金在质子及多束粒子下的辐照硬化是辐照偏析及可能的位错环硬化综合作用的结果。     

Neutron irradiation hardening and microstructure changes in Fe-Mn binary alloys

Irradiation hardening and microstructure changes in Fe-Mn binary alloys were investigated after neutron irradiation at 290 °C and up to 0.13 dpa. Significant irradiation hardening comparable to that of Fe-1 at.%Cu alloy was observed in Fe-1 at.%Mn alloy. Manganese increases the number density of dislocation loops, which contributed to the observed irradiation hardening. Manganese serves as a nucleus of the loop by trapping interstitial atoms and clusters, preventing 1D motion of the loops.     

Irradiation induced dislocation loop and its influence on the hardening behavior of Fe-Cr alloys by an Fe ion irradiation

Nano indentation analysis and transmission electron microscopy observation were performed to investigate a microstructural evolution and its influence on the hardening behavior in Fe-Cr alloys after an irradiation with 8 MeV Fe⁴⁺ ions at room temperature. Nano indentation analysis shows that an irradiation induced hardening is generated more considerably in the Fe-15Cr alloy than in the Fe-5Cr alloy by the ion irradiation. TEM observation reveals a significant population of the a₀<1 0 0> dislocation loops in the Fe-15Cr alloy and an agglomeration of the 1/2a₀<1 1 1> dislocation loops in the Fe-5Cr alloy. The results indicate that the a₀<1 0 0> dislocation loops will act as stronger obstacles to a dislocation motion than 1/2a₀<1 1 1> dislocation loops.     

A polycrystalline modeling of the mechanical behavior of neutron irradiated zirconium alloys

Zirconium alloys used as fuel cladding tubes in the nuclear industry undergo important changes after neutron irradiation in the microstructure as well as in the mechanical properties. However, the effects of the specific post-irradiation deformation mechanisms on the mechanical behavior are not clearly understood and modeled. Based on experimental results it is discussed that the kinematic strain hardening is increased by the plastic strain localization inside the dislocation channels as well as the only basal slip activation observed for specific mechanical tests. From this analysis, the first polycrystalline model is developed for irradiated zirconium alloys, taking into account the irradiation induced hardening, the intra-granular softening as well as the intra-granular kinematic strain hardening due to the plastic strain localization inside the channels. This physically based model reproduces the mechanical behavior in agreement with the slip systems observed. In addition, this model reproduces the Bauschinger effect observed during low cycle fatigue as well as the cyclic strain softening.     

H、He离子辐照对国产ZIRLO合金氧化膜的影响

针对国产ZIRLO合金开展了H、He离子辐照对其腐蚀性能影响的研究。对国产ZIRLO合金样品分别进行高温（300 ℃）H、He离子辐照试验,辐照峰值剂量为1 dpa,之后进行模拟一回路腐蚀试验。通过腐蚀增重方法得到腐蚀动力学曲线。利用慢正电子湮没多普勒展宽谱对未辐照样品和辐照样品进行微观结构表征,用透射电子显微镜对腐蚀125 d的样品进行微观结构表征。结果表明,H、He离子辐照并未改变ZIRLO合金的腐蚀机理。He离子辐照产生的空位团可促进腐蚀过程中裂纹形核,增加了氧扩散通道,减少氧扩散激活能,导致腐蚀初期有明显的加速效应。H离子辐照对腐蚀的加速现象不如He离子辐照明显,原因是H离子辐照产生H-空位复合缺陷对氧扩散激活能减少作用较小。     

He ion irradiation damage in Fe/W nanolayer films

We report on the evolution of microstructure and mechanical properties of Fe/W multilayers subjected to helium ion irradiations. Sputtered Fe/W multilayers with individual layer thickness, varying from 1 to 200 nm, were subjected to He⁺ ion irradiation with a peak displacement per atom value of 6 at ambient temperatures. Helium bubbles, 1-2 nm in diameter, were observed in Fe and W, and more so along layer interfaces. The magnitude of hardness variation after radiation depends on the individual layer thickness. Radiation hardening is observed in specimens with individual layer thickness of ?5 nm. At smaller layer thickness, the hardness barely changes. Analysis indicates that radiation hardening may originate mainly from dislocation loops and partially from He bubbles.     

Microstructural study of an irradiated high-nickel alloy by X-ray line profile analysis and TEM observations

The microstructure was investigated in previously deformed specimens of a high-nickel alloy with four levels of initial dislocation density, both before and after irradiation in BOR-60 reactor at 370–400°C and three displacement rates. The network dislocation density, dislocation loop radius and loop number density were determined by both X-ray line profile analysis and TEM observations. The dependence of dislocation structure parameters on irradiation conditions and initial network density was obtained.     

Isolating the effect of radiation-induced segregation in irradiation-assisted stress corrosion cracking of austenitic stainless steels

Post-irradiation annealing was used to help identify the role of radiation-induced segregation (RIS) in irradiation-assisted stress corrosion cracking (IASCC) by preferentially removing dislocation loop damage from proton-irradiated austenitic stainless steels while leaving the RIS of major and minor alloying elements largely unchanged. The goal of this study is to better understand the underlying mechanisms of IASCC. Simulations of post-irradiation annealing of RIS and dislocation loop microstructure predicted that dislocation loops would be removed preferentially over RIS due to both thermodynamic and kinetic considerations. To verify the simulation predictions, a series of post-irradiation annealing experiments were performed. Both a high purity 304L (HP-304L) and a commercial purity 304 (CP-304) stainless steel alloy were irradiated with 3.2 MeV protons at 360 °C to doses of 1.0 and 2.5 dpa. Following irradiation, post-irradiation anneals were performed at temperatures ranging from 400 to 650 °C for times between 45 and 90 min. Grain boundary composition was measured using scanning transmission electron microscopy with energy-dispersive spectrometry in both as-irradiated and annealed samples. The dislocation loop population and radiation-induced hardness were also measured in as-irradiated and annealed specimens. At all annealing temperatures above 500 °C, the hardness and dislocation densities decreased with increasing annealing time or temperature much faster than RIS. Annealing at 600 °C for 90 min removed virtually all dislocation loops while leaving RIS virtually unchanged. Cracking susceptibility in the CP-304 alloy was mitigated rapidly during post-irradiation annealing, faster than RIS, dislocation loop density or hardening. That the cracking susceptibility changed while the grain boundary chromium composition remained essentially unchanged indicates that Cr depletion is not the primary determinator for IASCC susceptibility. For the same reason, the visible dislocation microstructure and radiation-induced hardening are also not sufficient to cause IASCC alone.