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%。     

氦离子辐照对镍基合金硬度的影响

在室温下对镍基合金进行了氦离子辐照,利用纳米压痕仪测试了微观硬度,利用慢正电子多普勒展宽谱(Doppler Broadening Spectrum,DBS)和透射电子显微镜(Transmission Electron Microscope,TEM)分析了微观缺陷,利用离子束分析弹性反冲探测(Elastic Recoil Detection,ERD)技术测量了氦的浓度深度分布。结果显示合金样品的硬度随剂量而增大,退火后合金样品硬度增量有所减小,并观测到氦泡生成。合金硬化的主要原因是由于氦离子辐照产生了1-7 nm的缺陷团簇,而退火后不稳定缺陷的回复及氦-空位复合体数量的减少造成了硬化强度减弱。     

氦离子辐照烧结碳化硅损伤效应研究

利用中国科学院近代物理研究所320 kV高压平台提供的氦离子辐照烧结碳化硅,辐照温度从室温到1 000 ℃,辐照注量为1015~1017 cm-2。辐照完成后,进行退火处理,然后开展透射电子显微镜、拉曼光谱、纳米硬度和热导率测试。研究发现,烧结碳化硅中氦泡形核阈值注量低于单晶碳化硅。同时,氦泡形貌和尺寸与辐照温度、退火温度有关。另外,对辐照产生的晶格缺陷、元素偏析进行了研究。结果表明,辐照产生了大量的缺陷团簇,同时氦泡生长也会发射间隙子,在氦泡周围形成间隙型位错环。在晶界处,容易发生碳原子聚集。辐照导致材料先发生硬化而后发生软化,且热导率降低。     

高温He离子辐照引起的Hastelloy N合金微结构变化研究

本文采用30 keV的He离子辐照Hastelloy N合金,辐照温度为500C,剂量分别为:1×10¹⁵、×10¹⁵、1×10¹⁶He⁺/cm²。利用扫描电子显微镜（SEM）研究了辐照后块体样品的表面形貌。结果表明,辐照后的块体样品均观察到了表面起泡现象。利用透射电子显微镜（TEM）研究了辐照后TEM样品微结构的变化。结果表明,低剂量(1×10¹⁵ He⁺/cm²)辐照的样品中出现了黑斑缺陷;随着辐照剂量的增加,开始出现位错环及纳米级的氦泡,同时黑斑密度减少;当剂量增至1×10¹⁶He⁺/cm²,位错环以及氦泡的尺寸和密度明显增大,晶界处氦泡更加密集。     

锆合金原位离子辐照条件下的位错环演化研究

位错环演化是核用锆合金辐照组织演化的主要特征之一,对合金辐照后的力学性能（强度、塑性等）有着决定性的影响。目前,锆合金辐照位错环演化的实验研究主要基于离位中子或离子辐照,无法直接观察位错环的演化过程。为了更深入地理解锆合金辐照下的微观组织演化,本工作采用先进的原位离子辐照实验方法,实时观察Zr-2合金位错环的演化过程,揭示不同辐照损伤剂量和温度对演化过程的影响规律,并结合弥散障碍物硬化模型对合金的辐照硬化性能进行了评估,验证了原位离子辐照用于研究锆合金包壳材料辐照后位错环演化和力学性能评价的可行性和先进性。      

He+离子辐照Inconel 718合金表面形貌改变及机理研究

研究了3种不同剂量He⁺离子辐照后Inconel 718合金的形貌变化规律及其形成机理。结果表明,He⁺离子辐照会在合金表面形成纳米多孔结构,其孔径会随辐照剂量的增加而增大。此外,He⁺离子辐照还会破坏合金表面δ相并导致碳化物的持续溅射损耗,且这一现象会随着辐照剂量的增加而愈发严重。由于辐照过程中氦泡间微观应力σ n作用会引起毗邻材料断裂及氦泡合并长大,且辐照溅射作用又会导致氦泡上层薄膜的损耗甚至破裂,因而这也是He⁺离子辐照Inconel 718合金表面纳米多孔结构的形成机制。     

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。     

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。     

两类钒基合金的辐照硬化研究

钒合金(V-Cr-Ti系列)是重要的聚变堆结构候选材料,但是相比于铁素体/马氏体钢等其他候选材料,有关钒合金(V-Cr-Ti)的辐照损伤研究较为缺乏。利用载能离子束模拟聚变堆中子辐照条件,对V-4Cr-4Ti和V-5Cr-5Ti两种样品进行了载能He离子和重离子辐照实验。实验采用离子束梯度减能方法在样品中产生辐照损伤的坪区,利用纳米压痕技术测试材料的辐照硬化效应。结果表明,样品纳米硬度的深度递减现象可以用Nix-Gao模型很好描述,高能重离子辐照的样品中软基体效应可以有效避免;在He离子辐照情形,He浓度(以原子百万分率计(Atomic parts per million,APPM))/位移损伤(以每原子平均离位数计(Displacement per atom,DPA))大于4 200/0.2时,两种钒合金样品出现硬化饱和现象;相近位移损伤水平下,He与空位的结合导致缺陷集团的加速长大,致使材料的辐照硬化远大于重离子辐照情形。     

位错环上的Cr偏析对辐照硬化影响的原子尺度模拟研究

低温辐照脆化是影响低活化铁素体/马氏体（RAFM）钢服役的主要问题之一。RAFM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、析出物等）阻碍位错运动。本文利用分子动力学方法研究了bcc-Fe内刃型位错线与1/2〈111〉间隙位错环的相互作用,并对比分析了Cr偏析在位错环上对其硬化的影响。研究结果表明：刃型位错线挣脱位错环所需临界剪切应力（CRSS）与位错环的伯格斯矢量有关;在本文所研究条件下,在一定温度范围内,Cr偏析在位错环上会使得位错线挣脱所需CRSS增加,引起硬化增强。     

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。     

Influence of He ion irradiation on the microstructure and hardness of Ni-TiCNP composites

In the present study,samples of a titanium car-bide nanoparticle-reinforced nickel alloy(Ni-TiCNP com-posite)were irradiated with 1 MeV He ions at 700℃.The evolution of He bubbles and nanohardness was character-ized using transmission electron microscopy(TEM)and nanoindentation,respectively.TEM images showed that the size and number density of He bubbles in the grains were affected by the He ion fluence.The number density first increased significantly and then decreased with increasing ion dose,while the size exhibited an inverse trend.Moreover,the swelling induced by He bubbles continuously increased with increasing ion dose.He bub-bles also formed in the grain boundaries,interior of the TiC nanoparticles,and interfaces between the TiC nanoparticles and Ni matrix.Nanoindentation measurements indicated a decrease in nanohardness after irradiation,which is attributed to the disappearance of intrinsic dislocation lines caused by He ion irradiation.     

Radiation stability of ZrN under 2.6 MeV proton irradiation

Zirconium nitride is a promising alternative material for the use as an inert matrix for transuranic fuel, but the knowledge of the radiation tolerance of ZrN is very limited. We have studied the radiation stability of ZrN using a 2.6 MeV proton beam at 800 °C. The irradiated microstructure and hardening were investigated and compared with annealed samples. A high density of nano-sized defects was observed in samples irradiated to doses of 0.35 and 0.75 dpa. Some defects were identified as vacancy-type pyramidal dislocation loops using lattice resolution imaging and Fourier-filter image processing. A very slight lattice expansion was noted for the sample with a dose of 0.75 dpa. Hardening effects were found for samples irradiated to both 0.35 and 0.75 dpa using Knoop indentation.     

Irradiation damage in proton irradiated palladium-iron solid solutions

Transmission electron microscopy was used to investigate the irradiation damage, in particular irradiation induced precipitation (IIP), in Pd-base alloys containing 2, 8, 12 and 18 at % Fe. The specimens were irradiated mainly using 400 keV protons at a current density of 0.16 μA/mm² over the temperature range 110 to 750°C. A few samples containing 2 and 8% Fe were also irradiated using 3 MeV NiP⁺ ions. The irradiation microstructure of the proton irradiated alloys consists of dislocation loops over the temperature range 110 to 550°C and voids up to 650°C in all the alloys. IIP of Pd₃Fe was observed only in the Pd-18% Fe alloy between 110 and 500°C, irradiated to a dose of 0.9 dpa. Pd₃Fe was associated with dislocation loops, voids and grain boundaries. IIP was not observed in the Pd-2,8 and 12% Fe alloys proton irradiated to the same dose, nor to a higher dose of 1.5 dpa. It was also not observed in the 2 and 8% Fe alloys irradiated at 600 and 700°C by 3 MeV Ni⁺ ions.The absence of IIP in the more dilute alloys is attributed to the fast back diffusion of Fe atoms, which is due to the high mobility of vacancies in these alloys. This causes the Fe concentration at the sinks to remain below the solubility limit. Therefore, even though Fe is an undersized solute, the size effect alone is not sufficient for the production of IIP at point defect sinks in most Pd-Fe alloys. It is proposed that IIP can occur only when the alloy concentration is high enough to minimize the rate of back diffusion, which depends not only on the vacancy mobility but also on the concentration gradient near point defect sinks.     

Effect of Neutron Irradiation on Mechanical Properties of Al-Li Alloys

An increase in yield stress at room temperature was observed in Al-0.6W/0 Li alloy irradiated to thermal neutron doses of 2.9 × 10¹⁹ to 7.2 × 10¹⁹ cm^?2. The hardening of as-irradiated specimens is accompanied with yield point followed by jerky yield-elongation in the stress-strain curve. The radiation hardening could not be annealed out by heating for 30 min at temperatures up to 350°C, whereas the yield-elongation disappeared gradually with increasing heating temperature in the l mm diam. specimens; with the 2 mm diam. specimens the yield-elongation still remained even after post-irradiation heating for 30 min at 350°C. Strengthening accompanied by jerky yield-elongation is considered to be due to He atom clusters precipitated along the dislocation. The hardening observed in the specimens heat-treated after irradiation at temperatures above 250°C is caused by randomly distributed gas bubbles.

In heavily cold-worked Al-0.6%W/o Li specimens, recovery of work hardening occurred during neutron irradiation to 4.2 × 10¹⁹ cm^?2. Hardening due to gas bubbles was also observed in the cold-worked specimens. In Al-2.7W/0 Li alloy, an increase in yield stress took place in the specimens irradiated to 4.2 × 10¹⁹ cm^?2 and heated for 30 min at temperatures of 155° to 260°C. The hardening is thought to be due to re-precipitation of β-phase resolved during the neutron irradiation.     

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.     

The role of irradiated microstructure in the localized deformation of austenitic stainless steels

Localized deformation has emerged as a potential factor in irradiation-assisted stress corrosion cracking of austenitic stainless steels in LWR environments and the irradiated microstructure may be a critical factor in controlling the degree of localized deformation. Seven austenitic alloys with various compositions were irradiated using 2-3 MeV protons to doses of 1 and 5 dpa at 360 °C. The irradiated microstructure consisting of dislocation loops and voids was characterized using transmission electron microscopy. The degree of localized deformation was characterized using atomic force microscopy on the deformed samples after conducting constant extension rate tension tests to 1% and 3% strain in argon. Localized deformation was found to be dependent on the irradiated microstructure and to correlate with hardening originating from dislocation loops. Dislocation loops enhance the formation of dislocation channels and localize deformation into existing channels. On the contrast, voids mitigate the degree of localized deformation. The degree of localized deformation decreases with SFE with the exception of alloy B. Localized deformation was found to have similar dependence on SFE as loop density suggesting that SFE affects localized deformation by altering irradiated microstructure.