Journal of Nuclear Materials最新文献_第7页

Development of an advanced hydride reorientation model for Zircaloy cladding and its experimental validation 为锆合金包层开发先进的氢化物重新定向模型及其实验验证

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-06 DOI: 10.1016/j.jnucmat.2024.155445

Changhyun Jo, Dahyeon Woo, Youho Lee

Hydride reorientation, which occurs under hoop stress during cooling, stands out as a primary mechanism for material degradation in spent fuel management. The radial hydride fraction (RHF) is strongly involved in the mechanical integrity of cladding, highlighting the necessity for a robust modeling framework for quantitative analysis. However, the predictability of previous thermodynamic models for hydride reorientation in reactor-grade Cold Worked Stress Relieved (CWSR) Zircaloy has been hindered due to the intricate nature of hydride reorientation and the difficulties in characterizing microstructures. Recent successful EBSD characterization of reactor-grade CWSR Zircaloy has revealed valuable insights into microstructural characteristics of hydrides, enabling advancements in the modeling framework of hydride reorientation. This study aims to develop a thermodynamic model specifically focused on predicting the RHF. The developed thermodynamic model, based on classical nucleation theory, integrates aforementioned microstructural findings, combined with the Hydride-Nucleation-Growth-Dissolution (HNGD) model to capture transient precipitation behavior during cooling. Extensive experimental validations demonstrate enhanced predictability of the model. Additionally, the study examines the sensitivities of hydride reorientation to hydrogen concentration, applied stress, and cooling rate. It also provides predictions on reorientation behavior for engineering implications such as extension of wet storage, matrix hardening, recrystallization, and thermal cycling, supported by plausible explanations rooted in the underlying physical mechanisms elucidated through the model.

氢化物在冷却过程中的环向应力作用下发生重新定向，是乏燃料管理中材料降解的主要机制。径向氢化物分量（RHF）与包层的机械完整性密切相关，因此需要一个强大的建模框架来进行定量分析。然而，由于氢化物重新定向的复杂性和微结构表征的困难性，以往反应堆级冷作去应力（CWSR）锆合金中氢化物重新定向的热力学模型的可预测性受到了阻碍。最近对反应器级 CWSR 锆合金成功进行的 EBSD 表征揭示了氢化物微观结构特征的宝贵见解，从而推动了氢化物重新定向建模框架的发展。本研究旨在开发一个专门用于预测氢化物再取向的热力学模型。所开发的热力学模型以经典成核理论为基础，将上述微观结构发现与氢化物成核-生长-溶解（HNGD）模型相结合，以捕捉冷却过程中的瞬态沉淀行为。广泛的实验验证表明，该模型的可预测性得到了增强。此外，研究还考察了氢化物重新定向对氢浓度、外加应力和冷却速率的敏感性。研究还预测了重新取向行为对工程的影响，如延长湿储存、基体硬化、再结晶和热循环，并根据模型阐明的基本物理机制提供了合理的解释。

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引用次数: 0

Displacement cascade bombardment of delta-hydrides in alpha-zirconium α-锆中δ-酸酐的置换级联轰击

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-06 DOI: 10.1016/j.jnucmat.2024.155446

Jose F. March-Rico, Richard W. Smith, Brendan M. Ensor

One of the factors affecting the in-pile performance of Zr-based alloys is the precipitation of hydrides once H concentrations exceed the terminal solubility limit. H transport and hydride precipitation/dissolution is commonly modeled in codes such as BISON, but most of the experimental data supporting these models has been collected on unirradiated materials. As such, there is considerable uncertainty as to the influence of irradiation effects. In this work, molecular dynamics simulations of displacement cascades were performed on δ-hydrides to elucidate: 1) the extent of H dissolution following cascade impacts and 2) any alterations to defect production characteristics when compared to cascades in bulk Zr. The immediate amount of H dissolved in a high-energy cascade impact is notable, but a considerable fraction of the dissolved H atoms are rapidly re-absorbed into the hydride at reactor-relevant temperatures. The amount of dissolved H also decreases with increasing hydride size. When considering the expected volume fractions of hydrides, it is not expected that the irradiation-induced H dissolution rate will significantly affect the availability of H in the Zr lattice. In terms of defect production, cascades which overlap δ-hydrides produced an order of magnitude more stable defects than equivalent-energy cascades in bulk Zr. Vacancy defects are predominantly contained within the hydride structure while interstitials clusters are found adjacent to the hydride surface. Interstitials are strongly repelled by the hydride structure which may drive the expulsion of cascade-generated interstitials to the hydride surface and impede athermal recombination. Thus, the interatomic potential used in this work predicted a significant alteration to the defect survival efficiency and a stark production bias in the availability of mobile defects in bulk Zr following hydride-overlapped displacement cascades.

影响 Zr 基合金桩内性能的因素之一是，一旦 H 浓度超过最终溶解极限，就会析出氢化物。H 传输和氢化物析出/溶解通常在 BISON 等代码中建模，但支持这些模型的大部分实验数据都是在未经过辐照的材料上收集的。因此，辐照效应的影响还存在相当大的不确定性。在这项工作中，对 δ-hydrides 进行了位移级联的分子动力学模拟，以阐明：1）级联撞击后 H 的溶解程度；2）与块状 Zr 中的级联相比，缺陷产生特征的任何改变。在高能级联撞击中立即溶解的 H 量是显著的，但相当一部分溶解的 H 原子在反应器相关温度下迅速被重新吸收到氢化物中。溶解的 H 量也会随着氢化物尺寸的增大而减少。考虑到氢化物的预期体积分数，预计辐照引起的 H 溶解率不会对 Zr 晶格中 H 的可用性产生重大影响。就产生缺陷而言，与块状锆中的等能级联相比，重叠δ-氢化物的级联产生的稳定缺陷要多出一个数量级。空位缺陷主要包含在氢化物结构中，而间隙团簇则出现在氢化物表面附近。间隙受到氢化物结构的强烈排斥，这可能会将级联产生的间隙驱逐到氢化物表面，并阻碍热重组。因此，这项工作中使用的原子间势预示着在氢化物重叠位移级联之后，块状锆中的缺陷存活效率会发生显著变化，移动缺陷的可用性也会出现明显的生产偏差。

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引用次数: 0

High-temperature structure, elasticity, and thermal expansion of ε-ZrH1.8 ε-ZrH1.8的高温结构、弹性和热膨胀

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-05 DOI: 10.1016/j.jnucmat.2024.155437

James R. Torres , Christopher A. Mizzi , Daniel A. Rehn , Tyler Smith , Scarlett Widgeon Paisner , Adrien J. Terricabras , Darren M. Parkison , Sven C. Vogel , Caitlin A. Kohnert , Mathew L. Hayne , Thomas J. Nizolek , M.A. Torrez , Tannor T.J. Munroe , Boris Maiorov , Tarik A. Saleh , Aditya P. Shivprasad

Zirconium hydride is a promising candidate material for nuclear microreactor applications as a solid-state moderator component, owing to its favorable neutronics properties and good thermal stability over other metal hydrides. In the present work, the crystal structure, thermal expansion, and elastic properties of the hydrogen-rich ε phase hydride were measured at elevated temperatures in the range 300–900 K. Samples were prepared by direct hydriding Zircaloy-4 metal – a nuclear-grade zirconium alloy. Room-temperature lattice parameters agree well with those reported from literature for unalloyed zirconium hydride and fall within an observed quadratic H-content dependence. The coefficients of thermal expansion, determined from lattice expansion and dilatometry, agree well within our work but were about 30 % lower than those reported by others for unalloyed hydrides. Density functional theory-based molecular dynamics simulations were used to compare with thermal expansion and elasticity measurements. Results showed lattice parameter temperature dependence and slope of thermal expansion align with those from measurements. Based on diffraction scans at select temperatures, ε phase remained stable in air up to at least 770 K. Likewise, dilatometry showed smooth thermal expansion up to the thermal decomposition temperature around 950 K. The precise decomposition temperature was not determined via diffraction due to sparse scanning. The complete elastic property measurements were gathered for ε-phase Ziracloy-4 hydride for the first time. Young's modulus was lower compared to the metal and δ hydride phases. High-temperature elasticity measurements were limited to <350 K due to acoustic dissipation effects.

与其他金属氢化物相比，氢化锆具有良好的中子特性和热稳定性，是一种很有希望作为固态慢化剂成分应用于核微堆的候选材料。本研究测量了富氢ε相氢化物在 300-900 K 高温范围内的晶体结构、热膨胀和弹性特性。室温下的晶格参数与文献中报道的非合金氢化锆的晶格参数非常吻合，并且属于观察到的二次氢含量依赖关系。根据晶格膨胀和膨胀率测定的热膨胀系数与我们的研究结果十分吻合，但比其他人报告的非合金氢化物的热膨胀系数低约 30%。基于密度泛函理论的分子动力学模拟与热膨胀和弹性测量结果进行了比较。结果表明，晶格参数的温度依赖性和热膨胀斜率与测量结果一致。根据选定温度下的衍射扫描，ε 相在空气中保持稳定，至少到 770 K。同样，扩张仪显示热膨胀平稳，直到 950 K 左右的热分解温度。首次收集到了ε相 Ziracloy-4 hydride 的完整弹性特性测量值。与金属相和δ氢化物相相比，杨氏模量较低。由于声耗散效应，高温弹性测量被限制在 350 K。

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引用次数: 0

On the theory of the nucleation of gas bubbles at grain boundaries and incoherent inclusions 关于在晶界和不连贯夹杂物上的气泡成核理论

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-05 DOI: 10.1016/j.jnucmat.2024.155443

M.S. Veshchunov

On the base of the critical analysis of two-dimensional models of the nucleation of gas filled bubbles at grain boundaries of helium-implanted specimens under the action of tensile stresses, a new model is developed within the framework of the Reiss theory of homogeneous nucleation in binary systems. This approach considers that gas bubbles are formed as a result of agglomeration in a binary system of vacancies and gas atoms at grain boundaries, avoiding significant simplifications of previous models based on the classical nucleation theory for single-component (unary) systems. The new model is extended to consider the nucleation of Xe bubbles at grain boundaries in UO₂ under irradiation conditions and can be used for numerical analysis of experimental observations after the foreseen implementation in a fuel performance code. A similar approach can be applied to the nucleation and growth of gas bubbles on incoherent inclusions, such as those observed in irradiated ODS steels.

在对拉伸应力作用下植入氦气的试样晶界上充满气体的气泡成核的二维模型进行批判性分析的基础上，在二元体系中均匀成核的雷斯理论框架内开发了一个新模型。这种方法认为气泡的形成是空位和气体原子在晶界的二元体系中聚集的结果，避免了以往基于经典成核理论的单组分（一元）体系模型的重大简化。新模型可扩展用于考虑辐照条件下二氧化钛晶界处 Xe 气泡的成核问题，并可在预期的燃料性能代码实施后用于实验观测的数值分析。类似的方法也可应用于非相干夹杂物上气泡的成核和生长，例如在辐照 ODS 钢中观察到的气泡。

引用次数: 0

Microstructural changes and irradiation hardening behavior of V-4Cr-4Ti alloys irradiated with He ions using flash-electropolishing 使用闪速电抛光法辐照 He 离子的 V-4Cr-4Ti 合金的微观结构变化和辐照硬化行为

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-05 DOI: 10.1016/j.jnucmat.2024.155438

Ken-ichi Fukumoto , Yichen Zou , Takuya Nagasaka , Ryoya Ishigami

The flash-electropolishing of focused ion beam samples for V-4Cr-4Ti alloys is established, and the microstructures of high-purity V-4Cr-4Ti alloys after He ion irradiation are examined by transmission electron microscopy from room temperature to 700 °C. The correlation between irradiation hardening behavior and microstructural changes is clarified. During room temperature irradiation, defect clusters are formed at shallow positions in the specimens and no He bubbles are observed at the damage peak position. In contrast, 500 and 700 °C, TiCON precipitates are predominantly formed and He bubbles and voids were formed at the damage peak position. The results of nanoindentation tests and a comparison of irradiation hardening by irradiation damage indicate that the obstacle barrier strength factorαof TiCON is 0.45 while that of the irradiation defect clusters irradiated at room temperature is 0.10. Irradiation damage in the He ion range extends toward the interior of the specimens with increasing irradiation temperature.

建立了 V-4Cr-4Ti 合金聚焦离子束样品的闪速电抛光方法，并通过透射电子显微镜观察了高纯度 V-4Cr-4Ti 合金在 He 离子辐照后从室温到 700 ℃ 的微观结构。阐明了辐照硬化行为与微观结构变化之间的相关性。在室温辐照期间，试样的浅层位置形成了缺陷簇，在损伤峰值位置没有观察到 He 气泡。相反，在 500 和 700 ℃ 时，主要形成了 TiCON 沉淀，并在损伤峰值位置形成了 He 气泡和空隙。纳米压痕测试结果和辐照损伤的辐照硬化比较表明，TiCON 的障碍强度因子α 为 0.45，而室温辐照缺陷簇的障碍强度因子α 为 0.10。随着辐照温度的升高，He 离子范围内的辐照损伤向试样内部扩展。

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引用次数: 0

The effect of deuterium on defect production in irradiated tungsten 氘对辐照钨中缺陷产生的影响

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-04 DOI: 10.1016/j.jnucmat.2024.155422

V. Lindblad , D.R. Mason , F. Granberg

For fusion test reactors and power plants, one significant concern is the retention of hydrogen isotopes in the wall materials. The build-up of the radioactive and scarce fuel isotope tritium is of special concern, but knowing the retention of the other isotopes, such as deuterium, is also important. Deuterium is known to affect the mechanical properties of the wall material and most experiments are carried out on deuterium retention as it is safer to use than tritium. In addition to affecting the mechanical properties of the wall material, deuterium retention has been observed to affect the defect accumulation in the material. In this study, we investigate the phenomena and mechanisms responsible for the greater defect accumulation observed in tungsten when deuterium is present during irradiation. This is achieved computationally, utilizing molecular dynamics simulations and appropriate analysis tools. We found that deuterium will affect both the primary defect production as well as the recombination rate of defects in irradiated tungsten.

对于聚变试验反应堆和发电厂来说，氢同位素在炉壁材料中的滞留是一个重大问题。放射性稀缺燃料同位素氚的积累尤其令人担忧，但了解其他同位素（如氘）的保留情况也很重要。众所周知，氘会影响炉壁材料的机械性能，由于氘的使用比氚更安全，因此大多数实验都是针对氘的保留进行的。除了影响壁材的机械性能外，氘保留还被观察到会影响材料中的缺陷积累。在本研究中，我们研究了当氘存在于辐照过程中时，在钨中观察到的更大缺陷积累的现象和机制。我们利用分子动力学模拟和适当的分析工具进行了计算。我们发现，氘既会影响辐照钨中初级缺陷的产生，也会影响缺陷的重组率。

引用次数: 0

Crevice corrosion behavior of 316L austenitic steel in static liquid lead-bismuth eutectic at 550°C 550°C 静态液态铅铋共晶中 316L 奥氏体钢的缝隙腐蚀行为

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-02 DOI: 10.1016/j.jnucmat.2024.155441

Yuji Huang , Fanqiang Meng , Jiajian Shi , Lijun Zhang , Rui Yuan , Yingxue Chen , Feifei Zhang

Crevice corrosion is a significant form of localized corrosion that can lead to failures in structural components. However, there is a lack of research on this phenomenon in liquid lead-bismuth eutectic (LBE). In this study, the crevice corrosion behavior of 316 L exposed to stagnant LBE with the oxygen concentration of 1 × 10^–6 wt.% at 550 °C for 500 h was examined for the first time. Microstructural characterizations indicated 316 L is susceptible to crevice corrosion and reducing the crevices size facilitates a transition from oxidation to dissolution corrosion. Grain boundaries are able to provide more diffusion channels for oxygen, thereby promoting the development of Cr-rich oxides and chemical segregations beneath the surface scale. A simplified model elucidating the transportation of dissolved oxygen within the crevice environment of LBE has been developed.

缝隙腐蚀是一种重要的局部腐蚀形式，可导致结构部件失效。然而，目前还缺乏对液态铅铋共晶（LBE）中这一现象的研究。本研究首次考察了 316 L 在氧气浓度为 1 × 10-6 wt.%、温度为 550 °C 的液态铅铋共晶中暴露 500 小时的缝隙腐蚀行为。微观结构特征表明 316 L 易受缝隙腐蚀，缩小缝隙尺寸有利于从氧化腐蚀过渡到溶解腐蚀。晶界能够为氧气提供更多的扩散通道，从而促进富含铬的氧化物和表面鳞片下化学偏析的发展。我们建立了一个简化模型，以阐明溶解氧在枸橼酸钾盐缝隙环境中的迁移情况。

引用次数: 0

A finite element study on the irradiation-induced mechanical behaviors of aluminum-matrix radiation-shielding composites 铝基辐射屏蔽复合材料辐照诱导力学行为的有限元研究

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-02 DOI: 10.1016/j.jnucmat.2024.155440

Jiaqing Shi , Zheng Lv , Jian Wang , Wentao Tang , Yufei Liu , Zenglin Yang , Jian Yang , Zhimin Yang , Shuwang Ma

Aluminum-matrix radiation-shielding composites play a crucial role in advanced nuclear energy systems and fuel containers owing to their shielding design flexibility and desired structural compatibility. After being irradiated by neutrons, the shielding composites undergo irradiation damage and exhibit irradiation-induced mechanical effects such as irradiation hardening and embrittlement, which directly threaten the industrial application of the material. In this study, a finite element method was used to investigate the irradiation-induced mechanical behavior of radiation-shielding B₄C_P-W_P/Al composites. Using published data on the post-irradiation mechanical property evolutions of the matrix and shielding particles, and incorporating mechanisms of irradiation hardening and embrittlement, a finite element model was developed to describe the deformation of pristine and post-irradiation composites. Simulations of the post-irradiation mechanical properties of the aluminum-matrix radiation-shielding composites were conducted. The simulation results successfully reproduced the experimental findings for both the Al matrix and composites after irradiation. Furthermore, the stress-strain responses and deformation behaviors of the composites at different stages of irradiation damage are discussed. Finally, based on the simulation results, an artificial neural network was trained to efficiently predict the irradiation-induced mechanical behavior of the composites.

铝基辐射屏蔽复合材料因其屏蔽设计的灵活性和理想的结构兼容性，在先进核能系统和燃料容器中发挥着至关重要的作用。在受到中子辐照后，屏蔽复合材料会发生辐照损伤，表现出辐照诱导的力学效应，如辐照硬化和脆化，直接威胁到材料的工业应用。本研究采用有限元法研究了辐照诱导的辐照屏蔽 B4CP-WP/Al 复合材料的力学行为。利用已公布的基体和屏蔽粒子辐照后力学性能变化数据，并结合辐照硬化和脆化机理，建立了一个有限元模型来描述原始复合材料和辐照后复合材料的变形。对铝基辐照屏蔽复合材料辐照后的力学性能进行了模拟。模拟结果成功地再现了辐照后铝基体和复合材料的实验结果。此外，还讨论了复合材料在不同辐照损伤阶段的应力应变响应和变形行为。最后，根据模拟结果训练了一个人工神经网络，以有效预测复合材料在辐照诱导下的力学行为。

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引用次数: 0

TEM-EELS analysis reveals the W-atom mediated radiation-induced amorphization in M23C6 TEM-EELS 分析揭示了 M23C6 中由 W 原子介导的辐射诱导非晶化现象

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-02 DOI: 10.1016/j.jnucmat.2024.155439

Sho Kano , Huilong Yang , Masami Ando , Dai Hamaguchi , Takashi Nozawa , Hiroyasu Tanigawa , Kenta Yoshida , Tamaki Shibayama , Hiroaki Abe

To gain a mechanistic understanding of the phase stability of M₂₃C₆ upon irradiation, the bulk W-doped M₂₃C₆ (Cr-W-C system) in the range of 0–12 at.% W concentration was prepared and subjected to helium beam irradiation, following with a thorough electron energy loss spectroscopy (EELS) analysis. Radiation-induced amorphization (RIA) was observed only at the 4 W sample with a W concentration of ∼12 at.%. Analysis of the low-loss spectrum showed that the inelastic mean free path (λ) could be applied an effective indicator of the presence of an amorphous phase. The white line ratio of the carbon K-edge spectrum showed that the chemical bonding state in the crystalline state is mainly 2p^3/2 bonding, and it changes to dominantly 2p^1/2 bonding accompanying with the crystal-to-amorphous (c-a) transition. Discussion on the relationship between the change in λ (Δλ) and the lattice parameter (Δa) due to irradiation reveals that Δa is not dependent on Δλ, indicating that Δλ is mainly caused by the volume expansion due to the c-a transition. In addition, a crystalline state is remained even after a lattice parameter change of ∼1.5 % in 0 W and 1W-samples, whereas, a lattice expansion of ∼0.2 % would trigger the occurrence of crystal-to-amorphous transition in the 4W-sample. The detailed EELS analysis demonstrated that the constitutional W atoms play an important role in facilitating the occurrence of RIA in M₂₃C₆, that is, the phase instability accompanying the lattice expansion due to irradiation was emphasized by the addition of W in M₂₃C₆. The insights obtained here suggest that a higher W concentration in M₂₃C₆ is more susceptible to RIA, and therefore the resistance to amorphization is achievable by decreasing the W concentration in the steels.

为了从机理上了解 M23C6 在辐照下的相稳定性，制备了 W 浓度在 0-12 at.% 之间的掺 W M23C6（Cr-W-C 系统），并对其进行了氦束辐照，随后进行了全面的电子能量损失光谱（EELS）分析。仅在 W 浓度为 ∼12 at.% 的 4 W 样品中观察到辐射诱导的非晶化（RIA）。对低损耗光谱的分析表明，非弹性平均自由路径（λ）可作为非晶相存在的有效指标。碳 K 边光谱的白线比显示，晶体态的化学键状态主要是 2p3/2 键，随着晶体到非晶态（c-a）的转变，化学键状态变为主要是 2p1/2 键。讨论辐照引起的 λ 变化（Δλ）与晶格参数（Δa）之间的关系发现，Δa 与 Δλ 无关，这表明Δλ 主要是由 c-a 转变引起的体积膨胀造成的。此外，在 0W 和 1W 样品中，即使晶格参数变化了 ∼1.5 %，晶体状态仍然存在，而在 4W 样品中，晶格膨胀 ∼0.2 % 会引发晶体到非晶态的转变。详细的 EELS 分析表明，W 原子在促进 M23C6 中发生 RIA 的过程中发挥了重要作用，也就是说，M23C6 中添加 W 后，辐照引起的晶格膨胀所伴随的相不稳定性得到了强调。本文的研究结果表明，M23C6 中的 W 浓度越高，越容易发生 RIA，因此可以通过降低钢中的 W 浓度来抵抗非晶化。

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引用次数: 0

Influence of liquid lead and lead-bismuth eutectic on three alumina forming austenitic (AFA) steels through slow strain rate testing 通过慢应变速率测试，分析液态铅和铅铋共晶对三种氧化铝奥氏体钢（AFA）的影响

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2024-10-01 DOI: 10.1016/j.jnucmat.2024.155415

Christopher Petersson , Peter Szakalos , Rachel Pettersson , Mats Lundberg

Liquid metal embrittlement (LME) in three newly developed alumina-forming austenitic (AFA) alloys, two 50 kg batches and one 5-ton heat, was studied in the temperature range 350–600 °C in liquid Pb and 140–600 °C in LBE using slow strain rate testing (SSRT) in a low-oxygen environment. No significant decrease in the engineering strain was observed in either environment. However, the presence of secondary cracks along the length of the specimen and brittle intergranular areas on the fracture surfaces indicates that the AFA alloys do show a minor degree of embrittlement above 570 °C. This appears to be related to grain boundary wetting by Pb/LBE. At temperatures below 570 °C, this wetting effect does not seem to be strong enough to induce LME in the alloys, and their ability to form a sufficiently protective oxide means that they remain unaffected by LME. The results indicate that the AFA alloy group can perform sufficiently well in liquid Pb/LBE environments, and long-term testing should be carried out to determine their viability as candidate materials for use in Pb- and LBE-based cooling systems.

采用低氧环境下的慢应变速率试验（SSRT），研究了三种新开发的氧化铝奥氏体（AFA）合金（两批 50 千克和一批 5 吨热量）的液态金属脆性（LME）。在这两种环境中均未观察到工程应变的明显降低。然而，沿试样长度方向出现的二次裂纹和断裂面上的脆性晶间区域表明，AFA 合金在 570 °C 以上确实出现了轻微程度的脆化。这似乎与 Pb/LBE 的晶界润湿有关。在低于 570 °C 的温度下，这种润湿效应似乎不足以在合金中诱发 LME，而且合金能够形成足够的保护性氧化物，这意味着合金不会受到 LME 的影响。结果表明，AFA 合金组在液态 Pb/LBE 环境中的性能足够好，应进行长期测试，以确定它们是否可作为候选材料用于基于 Pb 和 LBE 的冷却系统。

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引用次数: 0