Journal of Nuclear Materials最新文献_第10页

Atomistic insights into the interfacial stability of graphene-reinforced Ni-based alloy composites after cumulative recoil events 累积反冲事件后石墨烯增强镍基合金复合材料界面稳定性的原子观察

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-09 DOI: 10.1016/j.jnucmat.2025.156371

Hai Huang , Xu Yu , Yanxin Jiang , Qing Peng , Guanyu Liu , Xiaobin Tang

Graphene (Gr)-reinforced metal matrix composites demonstrate excellent irradiation tolerance but face challenges in maintaining interfacial stability under extreme conditions. Using atomistic simulations, this study examines the evolution of the Gr/Ni-based alloy interface under 1000 cumulative recoils (∼0.333 dpa). Early cascade collisions minimally affect interfacial atomic order, but prolonged irradiation induces significant structural changes. Solute atoms progressively penetrate damaged Gr regions, thickening the interface. Gr retains portions of its six-membered ring structure and exhibits self-healing capabilities, balancing amorphous and crystalline phases even after extensive irradiation. Gr’s structural survival decays nonlinearly, stabilizing around 17.9 % after 1000 cascades. The damage evolution of Gr follows a four-stage progression characterized by distinct z-axis migration patterns influenced by solute atom interactions. Despite localized damage and disorder, Gr largely resists dissolution, maintaining its stabilizing role in interfacial integrity. Irradiation induces exponential decay of carbon-carbon bonds but growth of M–C bonds (where M denotes solute), paradoxically favoring metal-carbide formation over sp³ conversion. Furthermore, carbides nucleate preferentially at curled edges of Gr. These findings offer valuable insights into the irradiation-induced evolution of the composites for nuclear applications.

石墨烯（Gr）增强金属基复合材料具有优异的辐照耐受性，但在极端条件下保持界面稳定性面临挑战。利用原子模拟，本研究考察了Gr/ ni基合金界面在1000次累积后坐力（~ 0.333 dpa）下的演变。早期的级联碰撞对界面原子有序的影响很小，但长时间的辐照会引起明显的结构变化。溶质原子逐渐穿透受损的Gr区，使界面变厚。Gr保留了部分六元环结构，并表现出自愈能力，即使在广泛辐照后也能平衡无定形和结晶相。Gr的结构存活率呈非线性衰减，在1000级联后稳定在17.9%左右。在溶质原子相互作用的影响下，Gr的损伤演化具有明显的z轴迁移模式。尽管存在局部损伤和无序，但Gr在很大程度上抵抗溶解，保持了其在界面完整性中的稳定作用。辐照引起碳-碳键的指数衰减，但M -c键（M表示溶质）的增长，矛盾的是有利于金属碳化物的形成而不是sp3转化。此外，碳化物优先在Gr的卷曲边缘成核。这些发现为核应用复合材料的辐照诱导演化提供了有价值的见解。

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引用次数: 0

Morphology-governed rheology of RuO2 in borosilicate glass melts: Network formation and shear-thinning behavior 硼硅酸盐玻璃熔体中RuO2的形态控制流变性能：网络形成和剪切减薄行为

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-09 DOI: 10.1016/j.jnucmat.2025.156365

Xilei Duan , Qiang Zhang , Xueyang Liu , Zhenghua Qian , Kui Zhang , Guanyu Zhu , Yanbo Qiao

RuO₂ deposits during nuclear waste vitrification significantly alter the rheology of glass melts. This study systematically investigated the effects of RuO₂ content and crystal morphology on the rheological behavior of borosilicate glass melts using a high-temperature rotary viscometer. Acicular (RuO₂#a) and granular (RuO₂#g) crystals were prepared via a molten salt synthesis (MSS) method. The results demonstrate that increasing RuO₂ content markedly enhances melt viscosity and induces pronounced non-Newtonian behavior (shear-thinning). Crucially, the crystal morphology governs this effect: the high-aspect-ratio RuO₂#a facilitates the formation of a sample-spanning three-dimensional network, leading to a more significant viscosity increase and stronger shear-thinning compared to its granular counterpart (RuO₂#g) at an equivalent content. Fitting with the Cross model quantitatively confirms the superior network-forming ability of acicular crystals, yielding a significantly higher zero-shear viscosity (η₀) and a longer relaxation time (λ), which signifies a stronger and more stable agglomerate structure. This work establishes crystal morphology as a decisive factor in controlling the rheology of RuO₂-bearing glass melts.

核废料玻璃化过程中若o2的沉积显著改变了玻璃熔体的流变性。本研究采用高温旋转粘度计系统地研究了RuO2含量和晶体形态对硼硅酸盐玻璃熔体流变行为的影响。采用熔盐合成（MSS）法制备了针状（RuO2#a）和粒状（RuO2#g）晶体。结果表明，RuO2含量的增加显著提高了熔体粘度，并诱发了明显的非牛顿行为（剪切变薄）。至关重要的是，晶体形态决定了这种效应：高纵横比的RuO2#a促进了样本跨越三维网络的形成，与同等含量的颗粒状对应物（RuO2#g）相比，导致更显著的粘度增加和更强的剪切变薄。Cross模型的拟合定量证实了针状晶体具有较强的成网能力，具有较高的零剪切粘度（η0）和较长的弛豫时间（λ），表明其团聚体结构更强、更稳定。本研究确立了晶体形态是控制含氧化钌玻璃熔体流变性的决定性因素。

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引用次数: 0

High-temperature air oxidation of SDM-fabricated SS316L/Zr-alloy ATF cladding: Evidence for local ZrN–ZrO2 composite diffusion barriers sdm制造的SS316L/ zr合金ATF包层的高温空气氧化：局部ZrN-ZrO2复合扩散屏障的证据

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-08 DOI: 10.1016/j.jnucmat.2025.156366

Ji-Hyeok Choi , Ji Hyun Lee , Su Hyeong Kim , Unho Lee , Dong-Joo Kim , Young Soo Yoon

This study investigates the oxidation and nitridation of a double-layer SS316L/Zr-alloy accident-tolerant fuel cladding fabricated using a room-temperature, non-vacuum swaging–drawing methodology (SDM). Exposure to laboratory air up to 1,200°C produced Cr-rich and Fe-rich oxide regions at the interface, which served as an initial barrier. Additionally, local gaps inherent to the SDM process promoted the formation of a composite ZrN–ZrO₂ layer at the SS/Zr interface, which acted as a secondary diffusion barrier. Despite the breakdown of the external passive film and weld-induced cracking of the SS316L tube above 900°C, the double-layer cladding maintained its integrity longer than the bare Zr alloy. Mass-gain measurements and X-ray diffraction analysis revealed that the bare Zr-alloy cladding rapidly formed monoclinic ZrO₂ above 900°C, whereas the double-layer cladding preserved α/α′-Zr up to ≈1,100°C with delayed m/t-ZrO₂ formation. Raman peaks at ≈165, 224, and 503 cm⁻¹, combined with SAED data showing {111}(≈0.264 nm), {220}(≈0.162 nm) confirmed the presence of ZrN. The parabolic rate constant

k_{p}

(mg²·cm⁻⁴·s⁻¹) estimated at 600–1,200°C was consistently lower for the double-layer cladding than for the bare Zr-alloy cladding. Arrhenius fits indicated a higher apparent activation energy for the double-layer system, demonstrating the effectiveness of the strengthened diffusion barriers. These findings provide a mechanistic explanation for the delayed oxidation observed under open-air, loss-of-coolant-accident-like conditions and highlight the protective role of discontinuous yet functional ZrN–ZrO₂ composite layers.

本研究研究了采用室温非真空摆拔法（SDM）制备的双层SS316L/ zr合金耐事故燃料包壳的氧化和氮化过程。暴露在高达1200°C的实验室空气中，在界面处产生富cr和富fe的氧化物区域，作为初始屏障。此外，SDM过程固有的局部间隙促进了SS/Zr界面处复合ZrN-ZrO2层的形成，该层起到了二次扩散屏障的作用。当温度高于900℃时，SS316L管表面钝化膜发生破裂，焊致开裂，但双层熔覆层的完整性较裸Zr合金长。质量增益测量和x射线衍射分析表明，在900°C以上，裸zr合金熔覆层快速形成单斜ZrO2，而双层熔覆层保持α/α′-Zr高达约1100°C， m/t-ZrO2形成延迟。Raman峰在≈165、224和503 cm⁻¹处，结合SAED数据显示{111}（≈0.264 nm）、{220}(≈0.162 nm)，证实了ZrN的存在。抛物线速率常数kp （mg²·cm⁻⁴·s⁻¹）估计在600 - 1200°C之间，双层包覆层始终低于裸锆合金包覆层。Arrhenius拟合表明双层体系具有较高的表观活化能，表明强化扩散势垒的有效性。这些发现为在露天、类似冷却剂损失的条件下观察到的延迟氧化提供了机制解释，并强调了不连续但功能良好的ZrN-ZrO2复合层的保护作用。

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引用次数: 0

Ab-initio informed cluster dynamics simulation of self- and Xe diffusivity in uranium mononitride under irradiation 辐照下单氮化铀自扩散率和Xe扩散率的Ab-initio信息簇动力学模拟

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-07 DOI: 10.1016/j.jnucmat.2025.156360

Anton J. Schneider, Christopher Matthews, David A. Andersson, Michael W.D. Cooper

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuels considered for light water reactor and advanced reactor designs, as it presents significant advantages such as high uranium density (better economics) and high thermal conductivity and melting point (increased safety). Self- and fission gas diffusivities need to be better understood, given that they influence key fuel performance phenomena like swelling and fission gas release. Recently, radiation enhanced diffusivity was investigated in UN by means of cluster dynamics simulations relying on empirical potential-based parameterizations, the reliability of which highly depends on the interatomic potential accuracy. In this work, we refine this approach by determining, using ab-initio calculations, the properties of defect clusters containing vacancies, self-interstitials and Xe impurities. We also consider larger clusters than previous studies. The obtained dataset (formation enthalpies, entropies, and migration barriers) is used to parameterize a cluster dynamics model of mobile clusters, and to calculate the defect cluster concentrations under irradiation. This gives us access to the radiation enhanced self- and fission gas diffusivities. Although the resulting diffusivities are close to the values reported in the literature, we find important qualitative differences in the diffusion mechanisms. Capturing the correct mechanisms is crucial to properly describe the chemistry and fission rate dependence of the model.

单氮化铀（UN）是轻水反应堆和先进反应堆设计中考虑的氧化物燃料的陶瓷核燃料替代品之一，因为它具有高铀密度（更好的经济性）和高导热性和熔点（提高安全性）等显著优势。需要更好地理解自扩散和裂变气体扩散，因为它们影响关键的燃料性能现象，如膨胀和裂变气体释放。近年来，利用基于经验势参数化的簇动力学模拟方法研究了辐射增强扩散率，该方法的可靠性高度依赖于原子间势的精度。在这项工作中，我们通过使用从头算来确定含有空位、自间隙和Xe杂质的缺陷团簇的性质，从而改进了这种方法。我们还考虑了比以前的研究更大的集群。得到的数据集（形成焓、熵和迁移势垒）用于参数化可移动簇的簇动力学模型，并计算辐照下缺陷簇的浓度。这使我们能够接触到辐射增强的自我和裂变气体扩散系数。虽然得到的扩散系数接近文献中报道的值，但我们发现扩散机制存在重要的质的差异。捕获正确的机制对于正确描述模型的化学和裂变速率依赖性至关重要。

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引用次数: 0

In-situ observation of dislocation dynamics during dislocation channel broadening 位错通道展宽过程中位错动力学的原位观察

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-06 DOI: 10.1016/j.jnucmat.2025.156364

Jingfan Yang , Wei-Ying Chen , Xiaoyuan Lou

For the first time, this work revealed the dislocation dynamics during dislocation channel broadening through in-situ straining and radiation experiments in TEM. The study examines how the distribution of radiation-induced defects (mainly density and size) influences dislocation channel development. Serving as a follow-up study to our previous work (DOI: 10.1016/j.actamat.2024.119650), this paper provides direct evidence to visualize the root cause of dislocation channel broadening in additively manufactured (AM) 316L stainless steel (SS) after hot isostatic pressing (HIP) and compared to the wrought 316L SS. Radiation-induced defects were observed to pin the dislocation significantly, compared to the unirradiated condition, resulting in discontinued and segmented motion and constrained plastic flow through the dislocation channel. The distribution of radiation-induced defects in HIP AM 316L SS (smaller and denser defects) promoted more frequent out-of-plane cross-slip or double cross-slip, a key mechanism to form boarder dislocation channels than wrought counterpart.

本文首次通过原位应变和透射电镜辐射实验揭示了位错通道展宽过程中的位错动力学。研究了辐射缺陷的分布（主要是密度和尺寸）如何影响位错通道的发展。作为我们之前工作的后续研究(DOI：（10.1016/j.a actamat.2024.119650），本文提供了直接证据，可视化了增材制造（AM） 316L不锈钢（SS）在热等静压（HIP）后位错通道扩大的根本原因，并与变形的316L不锈钢（SS）进行了比较。与未辐照条件相比，观察到辐射引起的缺陷明显地钉住了位错，导致了中断和分割的运动，并限制了通过位错通道的塑性流动。与变形缺陷相比，辐射缺陷在HIP AM 316L SS中的分布更小、密度更大，促进了更频繁的面外交叉滑移或双交叉滑移，这是形成边界位错通道的关键机制。

引用次数: 0

LiOH-induced accelerated corrosion of Cr-coated Zr alloys: identification of Cr(OH)3 formation and enhanced degradation mechanisms at 1200°C 铬包覆Zr合金的lioh加速腐蚀：1200℃下Cr(OH)3形成和强化降解机制的鉴定

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-05 DOI: 10.1016/j.jnucmat.2025.156363

Chenghao Liu , Qing Li , Xuekui Qian , Saixiang Zhao , Jinbiao Bai , Jiakuan Chen , Yali Yang , Peng Song

The high-temperature steam oxidation behavior of Cr-coated Zr-4 alloys was investigated at 1200 °C under pure steam and lithiated steam (3500 ppm LiOH) conditions for up to 360 min to evaluate coating performance under accident-tolerant fuel cladding scenarios. Cr coatings with 15.6 μm thickness were deposited on Zr-4 substrates using multi-arc ion plating technology. Microstructural characterization using SEM, XRD, EDS, and EPMA revealed significant differences in oxidation mechanisms between the two environments. In pure steam, protective Cr₂O₃ layers formed with excellent substrate protection throughout the test duration, while LiOH-containing steam promoted dual-phase Cr₂O₃/Cr(OH)₃ formation with accelerated degradation. The presence of LiOH induced Cr(OH)₃ formation at oxide grain boundaries through Cr³⁺ + OH^- reactions, creating fast diffusion pathways for corrosive species. Multi-layered structures (Cr₂O₃/Cr/ZrCr₂/Zr substrate) developed with Zr-rich diffusion channels providing additional oxygen transport routes. XRD analysis confirmed substrate oxidation to ZrO₂ after 180–360 min in LiOH conditions, indicating coating failure, while pure steam maintained substrate integrity. The study elucidates the detrimental effect of LiOH on Cr coating protective capability and provides mechanistic understanding for accident-tolerant fuel cladding design under realistic primary circuit chemistry conditions.

在1200°C下，在纯蒸汽和锂化蒸汽（3500 ppm LiOH）条件下，研究了cr包覆Zr-4合金的高温蒸汽氧化行为，持续时间长达360分钟，以评估耐事故燃料包覆情景下的涂层性能。采用多弧离子镀技术在Zr-4衬底上沉积了厚度为15.6 μm的Cr涂层。通过SEM、XRD、EDS和EPMA等微观结构表征，揭示了两种环境中氧化机制的显著差异。在纯蒸汽中，Cr2O3保护层形成，在整个测试过程中具有良好的衬底保护作用，而含lioh蒸汽促进了Cr2O3/Cr(OH)3双相的形成，并加速了降解。LiOH的存在通过Cr3+ + OH-反应在氧化物晶界处形成Cr(OH)3，为腐蚀性物质创造了快速扩散途径。多层结构（Cr2O3/Cr/ZrCr2/Zr衬底）形成了富含Zr的扩散通道，提供了额外的氧运输途径。XRD分析证实，在LiOH条件下，180-360 min后，衬底氧化为ZrO2，表明涂层失效，而纯蒸汽保持了衬底的完整性。该研究阐明了LiOH对Cr包层防护能力的不利影响，为实际一次回路化学条件下的耐事故燃料包层设计提供了机理认识。

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引用次数: 0

First-principles calculations of positron annihilation parameters of fully-relaxed vacancies and vacancy-solute complexes in vanadium 钒中完全松弛空位和空位溶质配合物正电子湮灭参数的第一性原理计算

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-04 DOI: 10.1016/j.jnucmat.2025.156347

Sayani Mitra , S. Abhaya , R. Rajaraman

In this work, we have utilised self-consistent Two-Component DFT calculations for computing the positron lifetimes and Doppler Broadening Parameters of fully relaxed vacancies and vacancy-solute (Ti/Cr) complexes in Vanadium by including positron-induced forces using open-source code ABINIT. Inequal shifts of any particular atom associated with vacancy-like defects in the presence and absence of positron reveal the importance of the inclusion of positron-induced forces during the computation of positron annihilation parameters. An increase in positron lifetimes has been observed for fully relaxed vacancies compared to unrelaxed ones. The estimated binding energy of the Ti-vacancy complex is positive, and Ti shows inward movement during relaxation, which indicates a strong attraction of Ti in vacancy-solute complexes towards vacancy, whereas the negative binding energy of the Cr-vacancy complex and outward shift during atomic relaxation illustrate a very weak attraction in vacancy-solute complexes. Ti and Cr as solute atoms have been distinguished using the Ratio curves of the calculated momentum distribution of annihilating electron-positron radiation.

在这项工作中，我们利用自一致双分量DFT计算计算了钒中完全弛豫空位和空位溶质（Ti/Cr）配合物的正电子寿命和多普勒展宽参数，其中包括正电子诱导的力，使用开源代码ABINIT。在正电子存在和不存在的情况下，与类空位缺陷相关的任何特定原子的不等位移揭示了在计算正电子湮灭参数时包含正电子诱导力的重要性。与未放松的空位相比，已观察到完全放松空位的正电子寿命增加。Ti-空位配合物的结合能为正，Ti在弛豫过程中向内移动，表明Ti在空位-溶质配合物中对空位有很强的吸引力，而cr -空位配合物的结合能为负，原子弛豫过程中向外移动说明空位-溶质配合物对空位的吸引力很弱。利用计算得到的电子-正电子湮灭辐射动量分布的比值曲线，对Ti和Cr作为溶质原子进行了区分。

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引用次数: 0

The interaction of edge dislocations with hydrogen-helium bubbles in tungsten 钨中氢氦气泡与边缘位错的相互作用

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-04 DOI: 10.1016/j.jnucmat.2025.156362

Safqut Sanwar , Mathew M. Swisher , Cheng Sun

In fusion reactors, plasmas-facing components undergo degradation due to displacement damage and injection of gas impurities. Significant amounts of hydrogen (H) and helium (He) impurities can be introduced into materials through plasma exposure and nuclear transmutation, and their synergistic effects lead to the formation of mixed He/H bubbles under irradiation, changing their hardening mechanisms. In this study, molecular dynamics (MD) models were developed to study the interaction between ½<111> edge dislocations with mixed He/H-bubbles in W. The critical resolved shear stress (CRSS) that is required for dislocation breakaway increases with He/H-to-vacancy ratio at 600 K, indicating strong pinning effects of mixed He/H bubbles. However, as the temperature increases up to 1400 K, the mixed He/H bubbles become unstable as the H atoms are increasingly emitted from the mixed He/H bubbles and migrate into W matrix, which lowers bubble pressure and CRSS. Overall, these results highlight the synergistic effects of He and H on dislocation behavior in irradiated W during deformation.

在聚变反应堆中，面对等离子体的组件由于位移损坏和注入气体杂质而发生降解。通过等离子体暴露和核嬗变，可以将大量的氢(H)和氦（He）杂质引入材料中，它们的协同作用导致辐照下He/H混合气泡的形成，改变其硬化机制。在本研究中，建立了分子动力学（MD）模型来研究w中1 / 2 <；111>；边缘位错与混合He/H气泡之间的相互作用。在600 K时，位错脱离所需的临界分解剪切应力（CRSS）随着He/H /空位比的增加而增加，表明混合He/H气泡具有很强的钉住作用。然而，当温度升高到1400 K时，混合He/H气泡变得不稳定，因为H原子越来越多地从混合He/H气泡中发射出来并迁移到W基体中，从而降低了气泡压力和CRSS。总的来说，这些结果突出了He和H对辐照W变形过程中位错行为的协同效应。

引用次数: 0

Numerical and analytical studies on tungsten fuzz growth under elevated surface temperatures 升高表面温度下钨毛丝生长的数值与分析研究

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-03 DOI: 10.1016/j.jnucmat.2025.156361

C.R. Zu , S.Y. Dai , J.Y. Chen , K.R. Yang , X.H. Li , W.F. Liu , S.G. Liu , V. Shymanski

Fiber-like nanostructures called “fuzz” can form on a tungsten (W) surface under the irradiation of helium (He) ions. Experiments conducted on the Large Power-Materials Irradiation Experimental System (LP-MIES) have demonstrated that the thickness of the fuzz presents an upward tendency as the surface temperature increases. However, when the surface temperature continues to increase, the thickness will decline instead. To study this phenomenon, dedicated numerical modellings of W fuzz growth at diverse surface temperatures have been performed using the SURO-FUZZ code. The newly incorporated pseudo-potential model into SURO-FUZZ aims to describe the annealing process of fuzzy nanofibers, which further enables us to explore the effects of elevated surface temperature on the growth and annealing of W fuzz. The simulation results concerning the growth thickness of W fuzz are in reasonable agreement with the experimental data obtained from LP-MIES. In order to save the computational resource, an analytical formula regarding the growth of W fuzz under varying surface temperatures has been proposed by leveraging the growth and annealing rates modelled by SURO-FUZZ. This analytical formula takes into account explicitly the suppression of fuzz thickness driven by annealing, which gives a reasonable agreement with the LP-MIES measurements at high fluences. Therefore, when contrasted with the modellings of SURO-FUZZ, the analytical formula provides a valuable means to predict the growth of W fuzz under long-term irradiation.

在氦离子的照射下，被称为“绒毛”的纤维状纳米结构可以在钨(W)表面形成。在大功率材料辐照实验系统（LP-MIES）上进行的实验表明，随着表面温度的升高，模糊层的厚度呈上升趋势。然而，当表面温度继续升高时，厚度反而会下降。为了研究这一现象，使用SURO-FUZZ代码对不同表面温度下的W绒毛生长进行了专门的数值模拟。在SURO-FUZZ中引入的伪势模型旨在描述模糊纳米纤维的退火过程，从而进一步探索表面温度升高对W模糊纤维生长和退火的影响。仿真结果与LP-MIES的实验数据吻合较好。为了节省计算资源，利用SURO-FUZZ模型的生长速率和退火速率，提出了W fuzzy在不同表面温度下的生长解析公式。该分析公式明确考虑了退火对模糊厚度的抑制，在高影响下与LP-MIES的测量结果有合理的一致性。因此，与SURO-FUZZ模型相比，该分析公式为预测W模糊在长期辐照下的生长提供了有价值的手段。

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引用次数: 0

Atomistic investigation on the helium effect on intergranular crack propagation in iron 氦对铁晶间裂纹扩展影响的原子性研究

IF 3.2 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-12-03 DOI: 10.1016/j.jnucmat.2025.156359

Yongjie Sun , Yiyi Ma , Lei Peng, Shangming Chen, Jingyi Shi, Yeshang Hu

The helium bubbles induced by high-energy neutron irradiation can cause intergranular fractures in ferritic/martensitic steel. To elucidate the susceptibility of different grain boundaries (GBs) to helium-induced embrittlement, molecular dynamics method was applied to study the effect of helium bubbles on the intergranular crack propagation of five representative symmetric tilt grain boundaries (STGBs) in iron, during the tensile processes at 600 K. The GB models with sub-nanometer helium-vacancy (He-V) cluster distributions were derived from equilibrium helium evolution at a helium concentration of 2000 appm. The intergranular crack was inserted in the middle of the GB model to study the orientation effect on crack propagation. The results indicate that in clean GBs, the intergranular crack propagation along lower-energy boundaries exhibit bidirectional ductility, whereas crack propagation along higher-energy boundaries demonstrates unidirectional brittleness. However, He-V clusters can degrade the plastic behavior around the crack tip and promote the propagation of intergranular crack. Crack propagation resistance correlates with GB energy, with the coherent twin boundary Σ3{112} showing the highest resistance to crack propagation. These findings elucidate the fundamental mechanism of helium-induced grain boundary embrittlement at the atomic scale and offer insights for mitigating helium-induced grain boundary embrittlement from the perspective of GB engineering.

高能中子辐照诱导的氦泡可引起铁素体/马氏体钢的晶间断裂。为了阐明不同晶界对氦致脆的敏感性，采用分子动力学方法研究了氦泡对铁中5种具有代表性的对称倾斜晶界（stgb）在600 K拉伸过程中沿晶裂纹扩展的影响。在氦浓度为2000 appm时，通过平衡氦演化得到了具有亚纳米级氦空位（He-V）簇分布的GB模型。在GB模型中间插入晶间裂纹，研究取向对裂纹扩展的影响。结果表明，在洁净gb中，沿低能边界的裂纹扩展表现为双向延性，而沿高能边界的裂纹扩展表现为单向脆性。然而，He-V团簇会降低裂纹尖端周围的塑性行为，促进晶间裂纹的扩展。裂纹扩展阻力与GB能量相关，相干孪晶界Σ3{112}对裂纹扩展阻力最大。这些发现在原子尺度上阐明了氦致晶界脆化的基本机理，并从GB工程的角度为减轻氦致晶界脆化提供了新的思路。

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引用次数: 0