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。     

Ar~+辐照Hastelloy C276显微结构演化多尺度模拟研究及实验验证

采用分子动力学结合团簇动力学研究了Hastelloy C276Ni基合金在Ar+辐照(室温,约10dpa)下的显微结构演化机理,开发了多尺度模拟程序Radieff,利用Radieff模拟了在Ar+辐照下C276中间隙位错环和孔洞的形核、长大过程。在武汉大学串列加速器-离子注入机-透射电镜一体化联机装置上开展了115keV Ar+辐照C276验证实验,采用一体化联机透射电镜观察了辐照缺陷尺寸及形貌。不同辐照剂量下位错环尺寸模拟结果与实验结果吻合很好。     

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。     

辐照条件下空洞超点阵形成过程的相场模拟

钨(W)具有高熔点、高热导率和优异的抗辐照能力等优点,是未来聚变堆面向等离子体部件的重要候选材料。然而中子辐照后的纯W中会产生空洞超点阵,严重影响其服役性能。本文改进了辐照条件下纯W中空洞超点阵形成过程的相场模型,采用更合理的体系总自由能函数表达形式,且考虑了空间与时间上随机分布的辐照点缺陷的产生。模拟结果表明:辐照过程中,间隙原子的定向扩散及其与空位的相互作用是空洞超点阵形成的主要原因;间隙原子沿不同方向的定向扩散形成了不同类型的空洞点阵;点阵中空洞的排列会随模拟时间的延长变得有序,空洞尺寸也会变得基本一致,而空洞形状并非标准的圆形,模拟结果与实验结果基本一致。     

辐照条件下空洞超点阵形成过程的相场模拟

钨（W）具有高熔点、高热导率和优异的抗辐照能力等优点,是未来聚变堆面向等离子体部件的重要候选材料。然而中子辐照后的纯W中会产生空洞超点阵,严重影响其服役性能。本文改进了辐照条件下纯W中空洞超点阵形成过程的相场模型,采用更合理的体系总自由能函数表达形式,且考虑了空间与时间上随机分布的辐照点缺陷的产生。模拟结果表明：辐照过程中,间隙原子的定向扩散及其与空位的相互作用是空洞超点阵形成的主要原因;间隙原子沿不同方向的定向扩散形成了不同类型的空洞点阵;点阵中空洞的排列会随模拟时间的延长变得有序,空洞尺寸也会变得基本一致,而空洞形状并非标准的圆形,模拟结果与实验结果基本一致。     

辐照导致奥氏体不锈钢内位错环演化行为模拟

为了描述和预测奥氏体不锈钢材料在辐照条件下位错环的演化行为,本文基于平均场速率理论结合分子动力学方法建立辐照诱导奥氏体不锈钢内位错环演化的物理模型,模拟电子、中子辐照诱导奥氏体不锈钢内位错环的演化行为,模拟结果与实验结果吻合很好。在此基础上探究了中子辐照产生的存活缺陷对材料内位错环演化行为的影响,研究表明级联碰撞过程中的缺陷存活率、缺陷成团率以及四间隙原子团簇所占的团簇份额是影响位错环演化行为的主要参数,而双间隙原子团簇份额和三间隙原子团簇份额之间比例对位错环的演化没有影响。     

双束（H+/e-）辐照下氢对12Cr-ODS铁素体钢组织损伤影响研究

利用氢离子（H⁺）束和电子（e^-）束双束（H⁺/e^-）同时辐照用化学浸润法制备的新型12Cr-ODS铁素体钢,研究其辐照损伤效应及组织变化。实验结果表明：由于氧化物的钉扎,基体内保持低密度位错网络;辐照初期随辐照剂量的增加,缺陷团在位错线上及其周围形成,尺寸增加,密度不断增大,并形成间隙型位错环;不同温度下辐照均产生小尺寸高密度的空洞,随辐照剂量的增大,空洞长大速度降低,空洞密度缓慢减小;不同温度下,辐照剂量达15dpa时,空洞肿胀均小于0.15%。对辐照产生的点缺陷与氢相互作用进行理论分析,12Cr-ODS铁素体钢在623~823K经双束辐照后,表现出良好的抗辐照损伤性。     

双束同时辐照低活性Fe-Cr-Mn(W、V)合金微观组织损伤的影响

研究了时效热处理低活性Fe Cr Mn(W、V)钢双束同时辐照损伤行为 ,结果表明 :92 3K/ 3 0 0 0h时效合金 ,经单独电子辐照 (1 0a- 1)出现低密度空洞 ,而经双束同时辐照的时效合金 ,在辐照初期就形成间隙型位错环和微小空洞。与无时效合金相比 ,随时效温度增加 ,空洞尺寸、空洞密度和空洞肿胀量增大。随时效温度的提高碳化物析出数量增多 ,奥氏体中合金元素Cr、Mn、W、V降低 ,He的存在有效地促进空洞肿胀量增大。     

非均匀形核对辐照诱导钨内微结构演化行为影响的团簇动力学模拟

金属钨(W)及其合金作为未来聚变堆最具应用前景的面向等离子体结构材料(PFMs),其服役性能直接影响聚变堆长期服役的安全性,辐照诱导W及其合金内微结构演化导致的辐照脆化现象始终是限制其工程应用的关键因素。本文基于分子动力学计算结果,进一步完善了辐照诱导材料微结构演化行为的团簇动力学模型,采用更加完备的物理模型描述材料内辐照缺陷的产生行为,并进一步探讨了W基体内辐照缺陷产生过程对微结构演化行为的影响。模拟结果表明,高能初始离位原子(PKA)诱发级联碰撞直接产生的缺陷团簇是W内位错环、空洞演化中最重要的形核机制;非均匀形核所产生的间隙团簇的扩散行为对位错环的长大行为有重要影响,会导致位错环尺寸分布中出现亚尖峰与台阶状形貌。     

双束(He+/e-)辐照下氦对12Cr-ODS铁素体钢组织损伤影响研究

利用氦离子(He+)束和电子(e-)束双束同时辐照化学溶胶法制备的新型12Cr-ODS铁素体钢.实验结果表明,辐照初期,随着辐照剂量增加,点缺陷团(黑斑)在基体内形成,密度不断增大,尺寸长大缓慢,辐照剂量为0.8dpa时形成问隙型位错环.不同试验温度下,辐照均产生小尺寸高密度的空洞,随着辐照剂量增加,空洞尺寸长大缓慢,...     

Numerical evaluation of the dislocation loop bias

We have performed numerical calculation of the capture efficiency of a dislocation loop in a finite toroidal reservoir, which is a more appropriate model for a dislocation loop than a spherical or cylindrical reservoir adopted in the previous models. It allows a direct evaluation of the capture efficiency and the bias for a loop of any size with a full account of the stress field in the loop region of influence. It is shown that the loop bias depends on the loop size, dislocation density and the interstitial to vacancy dilatation ratio. With increasing loop size its bias decreases or increases to the straight dislocation bias value if the dislocation density is low or high, respectively. The bias difference of loops of different sizes is shown to be the reason of a coexistence of vacancy and interstitial loops under irradiation. In the conventional case of the dislocation bias for interstitials, interstitial loops are expected to grow to larger sizes than vacancy loops, while in a special case of dislocation bias for vacancies, the opposite tendency is expected.     

Investigation of pulsed irradiation effects on interstitial loop growth in metals

A computer program for the solution of non steady-state diffusion equations describing the evolution of point defects and interstitial dislocation loops during pulsed and continuous irradiation is developed. The equations take into account mutual recombination of point defects, defect migration to dislocation loops and line dislocations, and the existence of equilibrium thermal vacancies. It is shown that interstitial loops grow from 2 to 9 run in diameter due to the surplus flux of interstitials in the non steady-state regime (dynamic preference) at 573 K. At 873 K the dislocation loops begin to shrink owing to line tension forces. Comparison of interstitial loop and vacancy behaviour for pulsed and continuous irradiation at 573 and 873 K is performed. It is shown that at pulse duration 2 × 10⁻⁶ s and repetition rate 100 pulses/s, pulsing does not affect the interstitial loop behaviour.     

Formation and Growth Process of Dislocation Loops in Zircaloys under Electron Irradiation

We have investigated the formation and growth process of dislocation loops in Zircaloys (Zrys) under electron irradiation using a high voltage electron microscope (HVEM). Dislocation loops are of great importance to degradation phenomena in fuel claddings, such as irradiation growth and reduced ductility in light water reactors. TEM specimens of three kinds of Zircaloys (Zry-2, Zry-4 and improved Zry-2) were irradiated with 1MeV electrons at temperatures from 320 K to 970K in the HVEM. Interstitial-type dislocation loops with the Burgers vector b = 1/3(1120) were formed on the pyramidal or the prismatic planes at the beginning of irradiation. It was found that the nucleation and growth process of those loops follows the kinetics based on the &-interstitial model where di-interstitials act as the nuclei of interstitial dislocation loops. Based on this model, migration energies of interstitials and vacancies were determined to be 0.15–0.22eV and 1.0–1.2eV, respectively, from the irradiation temperature dependence of the density and the growth rate of loops.     

In situ transmission electron microscopy of electron-beam induced damage process in nuclear grade graphite

Atomic level processes involved in the swelling and crack-closing in nuclear grade graphite under electron irradiation have been observed in real-time using transmission electron microscopy. Noise-filtered lattice images show the formation of vacancy loops, interstitial loops and resulting dislocations with unprecedented clarity. The dislocation dipoles formed via vacancy loops were found to undergo climb resulting in extra basal planes. Concurrent EELS studies showed a reduction in the atomic density because of the breakage of hexagonal carbon rings. The formation of new basal planes via dislocation climb in addition to the bending/breaking of basal planes leads to swelling and closing of micro-cracks.     

注氘低活化马氏体钢在电子辐照下的缺陷行为

低活化铁素体/马氏体(RAFM)钢被视为国际热核聚变反应堆以及聚变反应堆的第1壁候选结构材料之一,很多国家均在研究不同的RAFM钢,中国低活化马氏体(CLAM)钢的研究亦正在进行。核聚变会产生氢、氦、氘及氚,这些气体元素与辐照缺陷结合在一起,对材料的辐照性能会产生较大影响。本文对注氘后不同温度下的辐照后微观结构进行研究。试验利用日本北海道大学的JEOL-1300高压电子显微镜研究注氘CLAM钢从室温到873K在1250keV电子辐照下的微观结构变化。研究结果表明,在电子辐照下,注氘产生的缺陷团会出现消失和长大两种现象,意味着间隙型与空位型位错环在注氘过程中同时产生。并研究了注氘产生的空洞。     

Mechanisms of radiation induced creep and growth

Irradiation creep occurs primarily because the applied stress causes the evolving microstructure to respond in an anisotropic fashion to the interstitial and vacancy fluxes. On the other hand, irradiation growth requires the response to be naturally anisotropic in the absence of applied stress. Four fundamental mechanisms of irradiation creep have been conjectured: stress induced preferred absorption (SIPA) of the point defects on the dislocations, stress induced preferred nucleation (SIPN) of point defects in planar aggregates (edge dislocation loops), stress induced climb and glide (SICG) of the dislocation network and stress induced gas driven interstitial deposition (SIGD). These mechanisms will be briefly outlined and commented upon. The contributions made by these mechanisms to the total strain are not, in general, mutually separable and also depend on the prevailing (and changing) microstructure during irradiation. The fundamental mechanism of irradiation growth will be discussed: it is believed to arise by the preferred condensation of point defects and climb of dislocation loops and network on certain crystallographic planes. The preferred absorption and nucleation is thus a consequence of natural crystallographic anisotropy and not due to any external stresses. Again the effectiveness of this mechanism depends on the prevailing microstructure in the material. In this connection attention will be particularly drawn to the significance of solute trapping, segregation at grain boundaries, dislocation bias for interstitials and transport parameters for an understanding of irradiation growth in materials like zirconium and its alloys; the relevance of recent simulation studies of growth in such materials using electrons to the growth under neutron irradiation will be discussed in detail and a consistent model of growth in these materials will be presented.     

Formation and annealing behavior of defect clusters in electron or He-ion irradiated Ti-rich Ti–Al alloys

In order to clarify the effect of He atoms on the formation and annealing behavior of defect clusters in Ti–Al alloys, a Ti–47 at.% Al intermetallic compound has been irradiated with electrons and He-ions. Helium-ion irradiation enhances the nucleation of defect clusters, especially of interstitial loops, at temperatures from 623 to 773 K in both γ-TiAl and ₂-Ti₃Al grains of the sample. However, there is little difference between the annealing temperature ranges of defect clusters in TiAl grains formed by He-ion or electron irradiation at 623 K. The dot-shaped clusters and interstitial loops grow scarcely during annealing, but are annihilated by annealing up to 923 K. Cavities are formed after irradiation with He-ions below 10 dpa at 773 K, but no cavities are formed by electron irradiation up to 30 dpa. The cavities in γ-TiAl and ₂-Ti₃Al grains survive after annealing even at 1053 K for 1.8 ks, keeping their density and diameter to be nearly the same as those in the as-irradiated grains.     

Positron lifetime calculations of defects in fusion irradiated beryllium

Numerical quantum-mechanical positron lifetime calculations were performed for mono-vacancies, di-vacancies, tri-vacancies and small nano-voids containing helium and hydrogen in neutron irradiated beryllium. Helium and hydrogen atoms in the sample after the irradiation are considered as atoms forming interstitial O-type loops. Spherical clusters of vacancies are included in the calculations as a reference. It was found that the presence of He and H atoms significantly changes the positron lifetime in irradiated beryllium. A correlation between the positron lifetime and mutual position of vacancies in nano-voids and interstitial loops was established.