表面效应对铁<100>间隙型位错环的影响

在材料辐照损伤过程中,间隙型位错环的形成及动力学行为严重影响材料在辐照条件下的服役行为.在常用的以体心立方铁为基的合金材料中,1/2<111>和<100>是两种主要的位错环,其对辐照损伤的影响一直都是核材料领域研究的热点之一.在之前的研究中,人们对{111}面与单个1/2<111>位错环的相互作用进行了深入研究,发现表面对位错环性质确实有重要的影响.采用分子动力学方法,在原子尺度详细研究了另一个重要的表面铁{100}面对<100>间隙型位错环动力学过程的影响.模拟发现位错环伯格斯矢量与表面法线方向的关系、距表面的深度、位错环之间的相互作用以及温度等,都对位错环与表面的相互作用产生重要影响,其中,表面作用下的伯格斯矢量的演化以及<100>位错环在此过程中的一维运动首次被发现.基于这些模拟结果,就<100>位错环对表面辐照损伤结构的影响进行详细地研究,给出<100>位错环对表面凹凸结构的贡献,这些结果为理解辐照过程中材料表面的演化提供一种可能的解释.     

注氢纯铝中间隙型位错环一维迁移现象的原位观察

核聚变堆材料在高能粒子辐照过程中会产生大量点缺陷,导致辐照脆性和辐照肿胀等现象.因而,研究点缺陷在辐照过程中的演变过程至关重要.点缺陷团簇的一维迁移现象是这种演变过程的主要研究内容之一.本文采用普通低压(200 kV)透射电镜,在室温条件下对注氢纯铝中的间隙型位错环在电子辐照下的一维迁移现象进行了观察和分析.在200 keV电子辐照下,注氢纯铝中的位错环可多个、同时发生一维迁移运动,也可单个、独立进行一维迁移运动.位错环沿柏氏矢量1/3<111>的方向可进行微米尺度的一维长程迁移,沿柏氏矢量1/2<110>的方向一维迁移也可达数百纳米.电子束辐照时产生的间隙原子浓度梯度是引起位错环一维迁移并决定其迁移方向的原因.位错环发生快速一维迁移时,其后会留下一条运动轨迹;位错环一维迁移的速率越快,运动的轨迹则越长,在完成迁移过后的几十秒内这些运动轨迹会逐渐消失.     

集中力作用下硅中位错结构

本文用化学侵蚀法研究了硅单晶样品在800—1000℃印压得到的位错“花结”。实验结果说明:印压产生的位错分布在{111}滑移面上;位错线的取向大部分是<110>或<112>方向。分析并观察到在压印下有两种位错环,一种是柏格斯矢量沿<110>方向并平行于(111)印压面;一种是柏格斯矢量沿<110>方向并与印压面相交。对位错环的结构进行了分析。     

铝镁合金中位错的运动与交滑移

本文研究了在电子显微镜的照明电子束作用下,铝镁合金中位错运动与交互作用的行为。螺型位错往往单个运动,并且很容易改变运动方向,产生多次双交叉滑移。滑移和交滑移首先在与膜面接近45°的{111}面上进行,位错的柏氏矢量为接近膜面的α/2<110>,这是与照明电子束所产生的应力与膜面平行一事相符的。运动着的位错可以通过其应变场激活近邻的位错,使之发生运动;亦可能受到其它位错的排斥作用而受阻或改变运动方向。     

退火直拉硅单晶中氧沉淀和诱生缺陷的电子显微镜研究

本文用超高压透射电子显微镜研究退火的高氧含量无位错直拉硅单晶中氧沉淀和诱生缺陷。在750—1050℃范围内氧沉淀是球状的α方英石。除了球状氧沉淀粒子之外还有一些具有{001}惯习面的方片状氧沉淀物。在950℃以上沿〈110〉方向从氧沉淀发射出冲压式棱柱位错环。这些位错环的柏氏矢量为α/2〈110〉、环面法线为〈110〉,它们是间隙型的位错环。这些位错环是从方片状氧沉淀物或从球伏氧沉淀粒子的聚集团发射出来的。当它们遇到障碍物时可能产生比较复杂的位错组态。实验中观察到由于层错攀移形成的台阶。热处理温度在850℃以下时,未观察到体内层错。 关键词：     

聚变材料钨辐照后退火形成的位错环特性及inside-outside衬度分析

对纯钨透射电镜薄膜样品在400℃进行了58 keV、1×10¹⁷ cm^-2(约0.1 dpa)的氘离子辐照,辐照后进行了900℃/1 h的退火处理.离子辐照产生了平均尺寸为(11.10±5.41)nm,体密度约为2.40×10²² m^-3的细小位错环组织,未观察到明显的空洞组织.辐照后退火造成了位错环尺寸的长大和体密度的下降,分别为(18.25±16.92) nm和1.19×10²² m^-3.通过透射电镜的衍射衬度分析,判断辐照后退火样品中的位错环主要为a/2<111>类型位错环.通过“一步法” inside-outside衬度分析判断位错环为间隙型位错环.辐照后退火还造成了较大位错环之间接触融合,形成不规则形状的大型位错环.此外,退火后样品中还观察到了尺寸为1—2 nm的细小空洞组织.     

bcc Fe中刃型位错的结构及能量学研究

基于位错理论，利用分子动力学方法建立了〈100〉{010}，〈100〉{011}，1/2〈111〉{011}和1/2〈111〉{112}刃型位错的芯结构，并计算了这四种刃型位错的形成能、位错芯能量和芯半径.计算结果表明：〈100〉{010}和〈100〉{011}刃型位错的形成能比1/2〈111〉{011}和1/2〈111〉{112}刃型位错的要高，这表明〈100〉刃型位错比1/2〈111〉刃型位错更难形成.而〈100〉{010}和〈100〉{011}刃型位错的芯半径比1/2〈111〉{011}和1/2〈111〉{112}刃型位错的小，这说明在1/2〈111〉刃型位错中位于奇异区的原子数多于〈100〉刃型位错，而这些原子要比完整晶体中的原子具有更大的活性.可见，1/2〈111〉刃型位错比〈100〉刃型位错更易运动，且〈100〉刃型位错在bcc Fe中难以形成. 关键词： bcc Fe 刃型位错 分子动力学模拟     

LiNbO_3晶体中包裹物和位错的X射线形貌研究

用X射线透射扫描形貌方法研究了LiNb0_3晶体中的包裹物和位错。在实验中发现了包裹物的相应于不同衍射矢量的X射线形貌与基于各向同性理论预言的形貌之间存在分歧,这被解释为弹性各向异性效应。同时还观察到Burgers矢量为最短点阵平移矢量1/3的纯刃型位错和次短点阵平移矢量1/3<0111>的纯螺型位错,以及由该螺型位错组成的纯扭转晶界。     

氦、氘对纯铁辐照缺陷的影响

在核聚变堆的辐照环境中, 核嬗变产物氢、氦对结构材料的抗辐照性能将产生很大的影响. 本实验采用离子注入和电子辐照模拟研究了氦和氘对具有体心立方结构的纯铁的影响. 采用离子加速器在室温分别对纯铁注入氦离子和氘离子, 经500℃时效1 h后在高压电镜下进行电子辐照.结果表明: 室温注氦和室温注氘的纯铁在500℃时效后分别形成间隙型位错环和空位型位错环. 在电子辐照下, 间隙型位错环吸收间隙原子而不断长大, 而空位型位错环吸收间隙原子不断缩小. 通过计算位错环的变化速率发现, 空位型位错环比间隙型位错环吸收了更多的间隙原子, 即室温注氘纯铁的位错偏压比室温注氦纯铁的偏压参量大, 这意味着相同实验条件下空位型位错环对辐照肿胀的贡献大于间隙型位错环对辐照肿胀的贡献. 利用氦-空位复合体和氘-空位复合体的结构, 分析了注氦和注氘后在纯铁中形成不同类型位错环的原因. 关键词： 氦 氘 辐照损伤 位错环     

磷酸铝(AlPO_4)晶体中缺陷的X射线形貌研究

用X射线衍射形貌法研究了AlPO_4晶体中的微观缺陷。在所研究的晶体中,主要晶体缺陷是生长层,沉淀物和位错。位错密度在晶体表面附近最大,晶体中部较低。位错主要起源于热应力和由沉淀物或生长层所造成的晶格畸变。多数位错的柏氏矢量是b=(a+c)<1123>型,部分的是b=a[2TT0]。分析了晶体缺陷与生长条件之间的关系。控制生长过程中的温度波动,特别是晶体出炉时的冷却速度,对提高晶体完美性是重要的。     

Microstructure and hardening effect of pure tungsten and ZrO2 strengthened tungsten under carbon ion irradiation at 700℃

Microstructure evolution and hardening effect of pure tungsten and W-1.5%ZrO₂ alloy under carbon ion irradiation are investigated by using transmission electron microscopy and nano-indentation. Carbon ion irradiation is performed at 700 ℃ with irradiation damages ranging from 0.25 dpa to 2.0 dpa. The results show that the irradiation defect clusters are mainly in the form of dislocation loop. The size and density of dislocation loops increase with irradiation damages intensifying. The W-1.5%ZrO₂ alloy has a smaller dislocation loop size than that of pure tungsten. It is proposed that the phase boundaries have the ability to absorb and annihilate defects and the addition of ZrO₂ phase improves the sink strength for irradiation defects. It is confirmed that the W-1.5%ZrO₂ alloy shows a smaller change in hardness than the pure tungsten after being irradiated. From the above results, we conclude that the addition of ZrO₂ into tungsten can significantly reduce the accumulation of irradiated defects and improve the irradiation resistance behaviors of the tungsten materials.     

超临界水冷堆燃料包壳材料的辐照损伤研究进展

超临界水冷堆（SCWR）是第四代核电站的主力堆型之一,高温、高压、超临界水环境下的辐照损伤问题是其燃料包壳材料面临的最大挑战。SCWR燃料包壳候选材料主要包括锆合金、奥氏体不锈钢、铁素体/马氏体不锈钢、镍基合金、ODS合金五大类,奥氏体不锈钢是最有希望的候选材料。介绍了近年来在这个领域国际上的主要研究进展。作者所在团队也对多种SCWR的候选材料进行了辐照损伤研究,包括:镍基合金C-276和718、铁素体/马氏体钢P92、奥氏体不锈钢AL-6XN和HR3C。对AL-6XN的氢离子辐照实验发现,辐照产生的缺陷主要是间隙型位错环,伯格斯矢量为1/3<111>,在较高剂量（5～7 dpa）辐照下,出现空洞肿胀。在氢滞留的影响下,位错环有着独特的演化规律,总结提出了位错环的四阶段演化过程。The Supercritical Water-cooled Reactor (SCWR) is one of the prior Generation IV advanced reactors. Irradiation damage is one of the key issues of fuel cladding materials which will suffer serious environment, such as high temperature, high pressure, high irradiation and supercritical water. The candidate materials contain zirconium alloys, austenitic stainless steels, ferritic/martensitic stainless steels, Ni-base alloys and ODS alloys. Austenitic stainless steels are the most promising materials. This paper summarized the international researches on irradiation effects in fuel cladding materials for SCWR. The group of authors also has done many researches in this field, including nickel-base alloy C-276 and 718, ferritic/martensitic steel P92 and austenitic stainless steel AL-6XN and HR3C. In AL-6XN austenitic stainless steels irradiated by hydrogen ions, dislocation loops were the dominant irradiation defects. At higher irradiation dose (5～7 dpa), the voids were found. All the dislocation loops were confirmed to be 1/3<111> interstitial type dislocation loops, and four evolution stages of dislocation loops with hydrogen retention were suggested.     

V-5Cr-5Ti合金弹性和力学性质的第一性原理研究

V-5Cr-5Ti合金作为核聚变堆第一包层的主要候选结构材料之一,但对其力学性质的理论研究相对较少.采用随机固溶体模型,利用第一性原理方法计算出V-5Cr-5Ti合金的弹性常数、体模量、剪切模量、杨氏模量、泊松比和柯西压力等,并与计算出的纯钒的相关数值进行对比,结果表明V-5Cr-5Ti合金具有良好的塑性和强度,但其塑性要略低于纯钒的.并对加入氧原子后的V-5Cr-5Ti合金进行了相关计算,通过对比计算结果发现,由于氧原子的加入,使V-5Cr-5Ti合金的塑性和强度都出现了不同程度的降低.最后对V-5Cr-5Ti合金和纯钒的理论强度进行了计算,并绘制出两者的应力-应变关系图,通过对比再次验证了上面的结论.     

Bubble-loop complexes induced by helium irradiation and effect of pre-irradiation and post-irradiation of hydrogen in reduced-activation ferritic/martensitic steel

In the present work, the synergistic effect of high concentration hydrogen and helium on the dislocation loops and bubbles as well as their correlations in reduced-activation ferritic/martensitic (RAFM) steels is investigated. Such an effect was transmuted from 14?MeV neutron irradiation and has been one of the most challenging issues for RAFM steels for future fusion reactors. After low dose (0.18?dpa) high concentration (5000 appm) single-ion helium irradiation at 723?K, very large dislocation loops were observed, and the majority of bubbles were inside dislocation loops, forming bubble-loop complexes. These bubble-loop complexes defects were also present in hydrogen/helium and helium/hydrogen sequential-ion irradiated steels. Pre-irradiated hydrogen ion effectively inhibited the later growth of loops induced by helium post-irradiation, and the higher the ratio of hydrogen to helium fluence, the greater the effect of inhibition. At high fluence of hydrogen pre-irradiation, the structure of bubble-loop complexes disappeared. On the other hand, hydrogen post-irradiation promoted the growth of loops induced by helium pre-irradiation, and the higher the ratio of hydrogen to helium fluence, the greater the effect of promotion. The mechanisms for hydrogen/helium synergistic effects are discussed.     

Hydrogen isotope accumulation in a vanadium alloy under different irradiations and the long-range interaction effect

Hydrogen and deuterium accumulation after irradiation of a vanadium alloy V-3.5 Ga with pulsed and steady-state plasma and hydrogen ions with a dose of 1.0 × 10¹⁹ cm^?2 has been studied. Hydrogen and deuterium saturation of surface layers 15–40 nm thick has been detected to occur both on the sides of irradiated and nonirradiated targets. Possible mechanisms of the observed effects are discussed.     

Effect of dose on irradiation-induced loop density and Burgers vector in ion-irradiated ferritic/martensitic steel HT9

Samples of F/M steel HT9 were irradiated to 20?dpa at 420°C, 440°C and 470°C in a transmission electron microscope with 1?MeV Kr ions so that the microstructure evolution could be followed in situ and characterised as a function of dose. Dynamic observations of irradiation-induced defect formation and evolution were made at the different temperatures. Irradiation-induced loops were characterised in terms of their Burgers vector, size and density as a function of dose and similar observations and trends were found at the three temperatures: (i) both a/2 <111> and a <100> loops are observed; (ii) in the early stage of irradiation, the density of irradiation-induced loops increases with dose (0–4?dpa) and then decreases at higher doses (above 4?dpa), (iii) the dislocation line density shows an inverse trend to the loop density with increasing dose: in the early stages of irradiation, the pre-existing dislocation lines are lost by climb to the surfaces while at higher doses (above 4?dpa), the build-up of new dislocation networks is observed along with the loss of the radiation-induced dislocation loops to dislocation networks; (iv) at higher doses, the decrease of number of loops affects more the a/2 <111> loop population; the possible loss mechanisms of the a/2 <111> loops are discussed. Also, the ratio of a <100> to a/2 <111> loops is found to be similar to cases of bulk irradiation of the same alloy using 5?MeV Fe²⁺ ions to similar doses of 20?dpa at similar temperatures.     

冷轧和退火对V-5Cr-5Ti合金组织结构和抗辐照硬化性能的影响

钒合金(V-5Cr-5Ti)是聚变堆第一壁以及包层的重要候选结构材料。不同加工工艺会对钒合金在聚变堆中的服役性能产生影响。本文利用兰州重离子研究装置(HIRFL)提供的337 MeV的高能Fe离子对不同程度冷轧(冷变形量分别为40%、60%和80%)以及冷轧后退火(1 273 K退火1 h)的V-5Cr-5Ti合金样品进行了辐照,研究了不同的冷轧和退火处理过程对材料抗辐照硬化性能的影响。电子背散射衍射技术(EBSD)测试结果显示,随着冷变形量的增加,样品中细小破碎晶粒比例增大,晶粒平均尺寸减小。退火处理后,细小破碎晶粒出现一定程度的长大,大晶粒几乎全部消失,晶粒尺寸分布更加均匀。维氏硬度结果表明随着冷变形量的增加,硬度随之增加,退火后硬度降低。辐照之后,材料硬度升高,出现了辐照硬化效应。在冷轧样品和退火样品中都观察到了辐照硬化效应随冷变形量的增加显著减弱的现象,这表明冷变形可以显著提高材料的抗辐照硬化能力。结合EBSD和硬度数据,对冷变形和退火处理引起钒合金抗辐照硬化性能变化的机理进行了讨论。讨论结果显示,冷轧使材料总的吸收尾闾增大,引起辐照硬化程度降低,退火处理使材料中晶界密度和位错密度降低,材料的总吸收尾闾降低,辐照硬化效应增加。     

6061-Al合金的自离子辐照损伤效应

铝合金是国内外研究堆的主要结构材料,在前期300#研究堆主要结构材料铝合金辐照性能研究的基础上,通过离子辐照研究6061-Al合金的微观结构损伤和引起的硬度变化,以开展较高辐照剂量下6061-Al合金损伤效应的前期探索。结果表明,经过自离子辐照后,6061-Al合金中产生了夹角为72°的位错环等缺陷,随着辐照剂量从0.218×1016 cm?2增加到4.367×1016 cm?2,缺陷密度明显增加,但选区电子衍射表明合金保持了很好的晶体结构,并没有发生非晶化。纳米压痕测试表明,不同辐照剂量下,样品中产生了不同程度的硬化,且微观硬度随着辐照剂量的增加而增加,当剂量增加到2.183×1016和4.367×1016 cm?2时,辐照硬化达到饱和,约为11%。研究结果可为初步预测较高中子辐照剂量下6061-Al合金结构和性能的变化提供数据支撑。     

Microstructure evolution of T91 steel after heavy ion irradiation at 550 ℃

Fe-Cr ferritic/martensitic (F/M) steels have been proposed as one of the candidate materials for the Generation IV nuclear technologies. In this study, a widely-used ferritic/martensitic steel, T91 steel, was irradiated by 196-MeV Kr⁺ ions at 550 ℃. To reveal the irradiation mechanism, the microstructure evolution of irradiated T91 steel was studied in details by transmission electron microscope (TEM). With increasing dose, the defects gradually changed from black dots to dislocation loops, and further to form dislocation walls near grain boundaries due to the production of a large number of dislocations. When many dislocation loops of primary a₀/2<111> type with high migration interacted with other defects or carbon atoms, it led to the production of dislocation segments and other dislocation loops of a₀<100> type. Lots of defects accumulated near grain boundaries in the irradiated area, especially in the high-dose area. The grain boundaries of martensite laths acted as important sinks of irradiation defects in T91. Elevated temperature facilitated the migration of defects, leading to the accumulation of defects near the grain boundaries of martensite laths.