Journal of Nuclear Materials最新文献_第9页

Investigating the evolution of U-10Mo fuel foil microstructures during multi-stage hot rolling using coupled potts model-finite element method simulations 利用波特斯模型-有限元法耦合模拟研究 U-10Mo 燃料箔在多级热轧过程中的微观结构演变

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

Journal of Nuclear Materials

Pub Date : 2024-09-24 DOI: 10.1016/j.jnucmat.2024.155427

William E. Frazier, Lei Li, Kyoo Sil Choi, Yucheng Fu, Zhijie Xu, Ayoub Soulami, Vineet V. Joshi

In this work, we study the microstructural evolution of U-10Mo foils over multiple stages of hot-rolling and reheating using our previously validated method coupling the Kinetic Monte Carlo Potts Model with the finite element method. Hot rolling and reheating refine the U-10Mo foil microstructure, but the relationships between the foil microstructure and recrystallization behavior over multiple successive reductions have complex relationships with the rolling schedule that have not yet been well quantified. Simulations were employed to forecast the impact of hot rolling reduction per pass, grain size variations, and uranium carbide (UC) distribution on the Johnson Mehl Avrami Kolmogorov (JMAK) recrystallization kinetics of the U-10Mo alloy, as well as it's post-rolling grain growth kinetics. Initial homogenized grain sizes varying from 100 µm to 1 mm and UC volume fractions ranging from 0 to 2 vol% were parametrically evaluated as a part of this study. While some of our simulation results support the findings of our previous analysis for conditions of single-pass rolling and annealing, our extended analysis shows that the grain size within the as-cast and homogenized foil can lead to significant changes in the in the distribution of strain within the final microstructure, which can slow grain coarsening over the final anneal. The magnitude of the hot rolling reduction per pass had a similarly strong impact on the distribution of strain within the foil microstructure and its subsequent grain growth behavior. The implications of these results on U-10Mo fuel foil fabrication procedures are discussed.

在这项工作中，我们使用之前通过验证的动力学蒙特卡洛波特斯模型与有限元方法相结合的方法，研究了 U-10Mo 箔在多个热轧和再加热阶段的微观结构演变。热轧和再加热完善了 U-10Mo 箔的微观结构，但是箔的微观结构与多次连续还原过程中的再结晶行为之间的关系与轧制进度之间存在复杂的关系，这些关系尚未得到很好的量化。模拟预测了每道次热轧减薄、晶粒大小变化和碳化铀（UC）分布对 U-10Mo 合金的 Johnson Mehl Avrami Kolmogorov（JMAK）再结晶动力学及其轧后晶粒生长动力学的影响。作为这项研究的一部分，我们对从 100 微米到 1 毫米不等的初始均化晶粒尺寸和从 0 到 2 Vol% 不等的 UC 体积分数进行了参数评估。虽然我们的一些模拟结果支持我们之前对单程轧制和退火条件的分析结果，但我们的扩展分析表明，铸造和均质化铝箔中的晶粒大小会导致最终微观结构中的应变分布发生显著变化，从而减缓最终退火过程中的晶粒粗化。每道工序的热轧减薄幅度对铝箔微观结构中的应变分布及其随后的晶粒生长行为也有类似的重大影响。本文讨论了这些结果对 U-10Mo 燃料箔制造程序的影响。

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引用次数: 0

Characterization of irradiation-induced dislocation loops in Vanadium at 25-500 °C 25-500 ℃辐照诱导的钒中位错环的表征

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

Journal of Nuclear Materials

Pub Date : 2024-09-24 DOI: 10.1016/j.jnucmat.2024.155428

Q. Zhang , L. Li , S. Chen , Y. Dong , E. Fu , X. Chang , L. Bao , X. Guo , K. Jin , Y. Xue

Vanadium-based alloys have emerged as promising candidates for structural materials in fusion applications. However, as its base metal, the response of V to irradiation has received limited attention in prior studies. To gain a fundamental understanding of the irradiation damage in V, its microstructure evolution under 6 MeV Ti ion irradiation at 25–500 °C is investigated in the present study, with the focus on the detailed and comprehensive characterization of the behavior of the irradiation-introduced dislocation loops. Under room temperature irradiation, the “black dot” dislocation loops agglomerate linearly into rafts, during which their Burgers vectors are well aligned. With the temperature increases to 300 °C, the size of rafts increases and the density decreases, while the size of small loops maintains similar to the room temperature irradiation condition. As the irradiation temperature reaches 500 °C, the defects become highly mobile, resulting in the formation of extended dislocation loops or lines with hundreds of nanometers in size, with the rafts vanishing. All the observable loops under this irradiation temperature range exhibit the Burgers vectors of a/2 < 111>. All the loops observed in the displacement region are identified to be interstitial-type, while a small portion of loops observed in the diffusion region under elevated temperatures are vacancy-type.

钒基合金已成为核聚变应用中很有前途的候选结构材料。然而，作为其基本金属，钒对辐照的响应在之前的研究中受到的关注有限。为了从根本上了解 V 的辐照损伤，本研究调查了其在 25-500 ℃ 的 6 MeV Ti 离子辐照下的微观结构演变，重点是对辐照引入的位错环的行为进行详细而全面的表征。在室温辐照下，"黑点 "位错环线性聚集成筏状，其布格斯矢量排列整齐。随着温度升高到 300 ℃，筏的尺寸增大，密度减小，而小环圈的尺寸保持与室温辐照条件下相似。当辐照温度达到 500 ℃ 时，缺陷变得高度流动，从而形成数百纳米大小的扩展位错环或线，筏消失。在此辐照温度范围内，所有可观察到的环都呈现出 a/2 <111>的伯格斯矢量。在位移区观察到的所有环路都被确定为间隙型环路，而在高温下的扩散区观察到的一小部分环路则为空位型环路。

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引用次数: 0

Microscopic characterisation of brittle fracture initiation in irradiated and thermally aged low-alloy steel welds of a decommissioned reactor pressure vessel 退役反应堆压力容器辐照和热老化低合金钢焊缝脆性断裂的显微特征分析

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

Journal of Nuclear Materials

Pub Date : 2024-09-21 DOI: 10.1016/j.jnucmat.2024.155423

N. Hytönen , S. Lindqvist , J. Lydman , Y. Ge , Z. Que , P. Efsing

Microstructure has a significant effect on material's integrity and in a heterogeneous weld microstructure the discontinuities affect the brittle fracture initiation and propagation and determine the fracture toughness. The knowledge of brittle fracture initiation mechanisms in high-Mn/high-Ni welds is limited. The brittle fracture initiation behaviour of the decommissioned Barsebäck Unit 2 reactor pressure vessel (RPV) welds of high-Mn/high-Ni weld metal from three different locations, the RPV head and the beltline regions, were investigated and compared with specimens from the surveillance program with high fluence. Systematic fractography has been performed on impact and fracture toughness specimens and the main features of the brittle fracture initiation in the component weld are presented and discussed. Two main types of initiators are identified as the weakest links to initiate the cleavage fracture and the initiation mechanism is found independent from the operation condition. The high-fluence surveillance specimens have a larger amount of intergranular cracking. The cleavage fracture initiation appears to be independent of the operation conditions but dependent on the welding process and metallurgical features. The findings aid in the development of improved material-property correlations which will result in better computational tools for predicting aging of welds based on microstructure.

微观结构对材料的完整性有重要影响，在异质焊接微观结构中，不连续性会影响脆性断裂的发生和扩展，并决定断裂韧性。对高锰/高镍焊缝中脆性断裂起始机制的了解十分有限。研究了退役的巴塞拜克 2 号机组反应堆压力容器（RPV）焊缝的脆性断裂起始行为，这些焊缝的高锰/高镍焊缝金属来自三个不同的位置，即 RPV 封头和腰线区域，并与高通量监视计划中的试样进行了比较。对冲击和断裂韧性试样进行了系统的断口分析，并介绍和讨论了组件焊缝中脆性断裂引发的主要特征。确定了两种主要类型的引发剂作为引发劈裂断裂的最薄弱环节，并发现引发机制与操作条件无关。高流变监控试样有较多的晶间裂纹。劈裂断口的引发似乎与操作条件无关，但取决于焊接工艺和冶金特征。这些发现有助于开发更好的材料-性能相关性，从而为根据微观结构预测焊缝老化提供更好的计算工具。

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引用次数: 0

Fuel performance code BERKUT-U to simulate the in-pile behavior of a single oxide or nitride fuel rod for fast reactors 燃料性能代码 BERKUT-U，用于模拟快堆中单根氧化物或氮化物燃料棒的堆内行为

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

Journal of Nuclear Materials

Pub Date : 2024-09-20 DOI: 10.1016/j.jnucmat.2024.155417

A.V. Boldyrev, A.P. Dolgodvorov, I.O. Dolinskiy, V.D. Ozrin, P.V. Polovnikov, V.E. Shestak, V.I. Tarasov, A.V. Zadorozhnyi

This paper describes the fuel performance code BERKUT-U, which the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) designed as part of the "Codes of The New Generation" subproject of the Proryv Project. The code aims to model oxide or nitride single fuel rod behavior with a gas or liquid metal sublayer under normal and accident conditions of a liquid metal-cooled fast reactor operation. The BERKUT-U code's models, incorporating the MFPR/R code, are grounded in the contemporary understanding of mechanisms governing the fundamental processes in fuel rods under irradiation, which substantially enhances the code's predictive ability in comparison with the engineering analogs. Simulations of the nitride and oxide fuel rod behavior in BN-600 and BOR-60 fast reactors have demonstrated good agreement with the post-irradiation examination data. Further validation is foreseen as the corresponding data are available.

本文介绍了俄罗斯科学院核安全研究所（IBRAE RAN）设计的燃料性能代码 BERKUT-U，该代码是 Proryv 项目 "新一代代码 "子项目的一部分。该代码旨在模拟在液态金属冷却快堆运行的正常和事故条件下，带有气态或液态金属下层的氧化物或氮化物单燃料棒的行为。BERKUT-U 代码的模型结合了 MFPR/R 代码，以当代对辐照下燃料棒基本过程的机制的理解为基础，与工程模拟相比，大大提高了代码的预测能力。对 BN-600 和 BOR-60 快堆中氮化物和氧化物燃料棒行为的模拟结果表明，与辐照后的检查数据非常吻合。随着相应数据的获得，预计将进行进一步验证。

引用次数: 0

Reducing the oxidation rate of Cr-coated Zr alloys under high temperature steam environment: An approach of an outer Zr coating 降低高温蒸汽环境下铬涂层 Zr 合金的氧化率：外层 Zr 涂层的方法

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

Journal of Nuclear Materials

Pub Date : 2024-09-20 DOI: 10.1016/j.jnucmat.2024.155421

Yanguang Cui , Man Zhang , Jianqiao Yang , Junkai Liu , Xintao Zhang , Qifeng Zeng , Junqiang Lu , Fen Zhao , Dayan Ma , Di Yun

Cr-coated Zr alloys are widely regarded as promising accident-tolerant fuel (ATF) cladding materials. However, the rapid consumption of the Cr coating through both oxidation and diffusion demands a thick coating, which is detrimental to neutron economic. In this study, a thin Zr coating was fabricated on the top of the Cr coating to act as a protective layer, aiming to reduce the oxidation consumption rate of Cr coatings. In-situ weight gain measurements were performed to determine the oxidation kinetics. The microstructural evolution of the coating/substrate interface and the oxide/coating interface were analyzed. The results show that the weight gain rate of the Zr-Cr coated Zr alloy sample is lower than that of the bare Cr-coated Zr alloy samples. An equiaxed grain structure composed of Cr/Zr mixed oxides was observed in the coating, which helps to inhibit elements diffusion and reduce the oxidation rate of Cr coatings. After steam oxidation at 1000 °C for 4 h, the sample structure consisted of the Zr substrate, a Cr₂Zr layer, a residual Cr coating, a Cr₂O₃ layer, a Cr/Zr mixed oxides layer, and a ZrO₂ layer.

人们普遍认为铬涂层锆合金是很有前途的事故耐受燃料（ATF）包层材料。然而，由于铬涂层在氧化和扩散过程中消耗很快，因此需要很厚的涂层，这不利于中子经济性。本研究在铬涂层的顶部制作了一层薄的锆涂层作为保护层，旨在降低铬涂层的氧化消耗率。为确定氧化动力学，进行了原位增重测量。分析了涂层/基体界面和氧化物/涂层界面的微观结构演变。结果表明，Zr-Cr 涂层 Zr 合金样品的增重率低于裸 Cr 涂层 Zr 合金样品。在涂层中观察到由 Cr/Zr 混合氧化物组成的等轴晶粒结构，这有助于抑制元素扩散并降低 Cr 涂层的氧化率。在 1000 ℃ 蒸汽氧化 4 小时后，样品结构由 Zr 基体、Cr2Zr 层、残余 Cr 涂层、Cr2O3 层、Cr/Zr 混合氧化物层和 ZrO2 层组成。

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引用次数: 0

Modeling of the damage and fracture behaviors of a SiC triplex tube during the burst test with elastomeric insert 碳化硅三联管在带弹性插件的爆破试验中的损伤和断裂行为建模

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

Journal of Nuclear Materials

Pub Date : 2024-09-20 DOI: 10.1016/j.jnucmat.2024.155420

Jinqiang Wang , Luning Chen , Zhiwei Lu , Guochen Ding , Qisen Ren , Jiaxiang Xue , Xiaobin Jian , Jing Zhang , Shurong Ding

The Silicon Carbide (SiC) triplex cladding tube has been regarded as one of the leading structures for the next-generation light water reactors, because of its larger safety margins under beyond-design basis transient conditions. In this study, a numerical simulation method is developed to reproduce the damage and fracture behaviors of a nuclear-grade SiC triplex cladding tube during the burst test. Especially, a three-dimensional continuum damage mechanics based (CDM-based) constitutive model is developed and validated for the SiC_f/SiC composites, with the predictions agreeing well with the experimental data under different loading conditions. By introducing cohesive surfaces in the monolithic layers of a SiC triplex tube, cracking of the monolithic layers and the subsequent local damage behaviors within the SiC_f/SiC composite layer are captured. The local tensile strength of ∼402 MPa is identified for the monolithic layers, corresponding to the first load drop during the burst test. The simulation results indicate that cracking of the monolithic layers leads to sharp increases in the locally enhanced hoop stresses and damage factors for the SiC_f/SiC composite layer, with slight influences on the field variables far away from the main crack; after the fast increase the evolution velocity of local damage factors slows down, reflecting the toughening effects of SiC_f/SiC composite. An assessment strategy for the gas leak tightness and structural integrity of the SiC triplex cladding during the accident sceneries is proposed to predict failure of the SiC_f/SiC composites with the critical damage factor, and it is necessary to simulate the damage and fracture behaviors in the multi-layer models with the cracking process of monolithic layers involved.

碳化硅（SiC）三联包壳管因其在超出设计基础的瞬态条件下具有更大的安全裕度而被视为下一代轻水反应堆的主要结构之一。本研究开发了一种数值模拟方法，用于再现核级 SiC 三重包壳管在爆裂试验中的损伤和断裂行为。特别是，针对 SiCf/SiC 复合材料开发并验证了基于连续损伤力学（CDM）的三维构成模型，其预测结果与不同加载条件下的实验数据十分吻合。通过在碳化硅三联管的单片层中引入内聚面，捕捉到了单片层的开裂以及随后碳化硅/碳化硅复合材料层内的局部损伤行为。确定了单片层的局部抗拉强度为 402 兆帕，与爆裂试验中的首次载荷下降相对应。模拟结果表明，整体层的开裂导致局部增强的环应力和 SiCf/SiC 复合材料层的损伤因子急剧增加，对远离主裂缝的场变量影响轻微；在快速增加后，局部损伤因子的演变速度减慢，反映了 SiCf/SiC 复合材料的增韧效应。提出了事故场景下 SiC 三重包层气体泄漏密封性和结构完整性的评估策略，以临界损伤因子预测 SiCf/SiC 复合材料的失效，有必要在多层模型中模拟损伤和断裂行为，并涉及单片层的开裂过程。

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引用次数: 0

The evolution of irradiation defects and hardening of CVD-SiC induced by He ions irradiation at 800°C 800°C He 离子辐照诱导的 CVD-SiC 辐照缺陷和硬化的演变

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

Journal of Nuclear Materials

Pub Date : 2024-09-19 DOI: 10.1016/j.jnucmat.2024.155419

Qiqi Li , Xiaoyue Li , Zhenbo Zhu , Xiangbin Ding , Min Liu

In this study, 500 keV He ions were used to irradiate CVD-SiC samples at 800 °C. The influence of doses on the microstructural evolution and hardness of the irradiated samples were investigated by TEM, Raman, and nanoindentation. TEM results show that He bubbles and dislocation loops appeared after irradiation, and their number densities increased with doses, resulting in the gradual decrease of the intensity of the TO peak in Raman spectra. Moreover, He platelets with strain field were observed in both stacking faults and matrixes. Nanoindentation results indicated that the irradiation hardening occurred, and the hardening degree was positively correlated with the irradiation dose.

本研究使用 500 keV He 离子在 800 °C 下辐照 CVD-SiC 样品。通过 TEM、拉曼和纳米压痕研究了剂量对辐照样品微观结构演变和硬度的影响。TEM 结果表明，辐照后出现了 He 气泡和位错环，其数量密度随剂量的增加而增加，导致拉曼光谱中 TO 峰的强度逐渐降低。此外，在堆叠断层和基体中都观察到了带有应变场的 He 小板。纳米压痕结果表明发生了辐照硬化，硬化程度与辐照剂量呈正相关。

引用次数: 0

Comparison of the corrosion behavior of four Fe-Cr-Ni austenitic alloys in supercritical water 四种铁-铬-镍奥氏体合金在超临界水中的腐蚀行为比较

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

Journal of Nuclear Materials

Pub Date : 2024-09-19 DOI: 10.1016/j.jnucmat.2024.155409

Tao Huang , Haozhan Su , Yuhao Zhou , Jiamei Wang , Lefu Zhang , Kai Chen

The corrosion behavior of four Fe-Cr-Ni austenitic alloys were investigated after exposure to deaerated supercritical water (SCW) at 650°C. Results show that the corrosion resistance follows the trend:Fe-29Cr-61Ni > Fe-16Cr-75Ni > Fe-25Cr-20Ni > Fe-21Cr-31Ni. The increase in Cr and Ni contents promote a transition from low-protective multilayer oxide scales to protective Cr₂O₃ scales. After surface grinding, the weight gains of all specimens dropped by one order of magnitude. Surface grinding induces the formation of ultrafine grains and a high density of dislocations, which enhances the corrosion resistance among all studied austenitic alloys. However, the beneficial effects of grinding diminish with increased Cr and Ni contents due to the pre-existing portion of Cr₂O₃ scales. The lower critical Cr content required to form Cr₂O₃ scales in high Ni austenitic alloys is primarily attributed to their lower solubility of O and the slower diffusion rates of Ni, which inhibit corrosion behavior.

研究了四种铁-铬-镍奥氏体合金在 650°C 下暴露于脱气超临界水 (SCW) 后的腐蚀行为。结果表明，耐腐蚀性的变化趋势为：Fe-29Cr-61Ni；Fe-16Cr-75Ni；Fe-25Cr-20Ni；Fe-21Cr-31Ni。铬和镍含量的增加促进了低保护性多层氧化鳞片向保护性 Cr2O3 鳞片的过渡。表面研磨后，所有试样的增重都下降了一个数量级。表面研磨促使形成超细晶粒和高密度位错，从而提高了所有研究奥氏体合金的耐腐蚀性。然而，随着铬和镍含量的增加，磨削的有益效果会减弱，原因是预先存在部分 Cr2O3 鳞片。在高镍奥氏体合金中，形成 Cr2O3 鳞片所需的临界铬含量较低，这主要是因为它们的 O 溶解度较低，镍的扩散速度较慢，从而抑制了腐蚀行为。

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引用次数: 0

A first-principles simulation study on solubility of La, Nd, Zr and Mo in UO2 and U3O8 关于镭、钕、锆和钼在二氧化铀和八氧化三铀中溶解度的第一原理模拟研究

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

Journal of Nuclear Materials

Pub Date : 2024-09-19 DOI: 10.1016/j.jnucmat.2024.155414

Zhiyuan Huang , Lidong Ma , Jianbao Zhang , Qing Zhou , Lei Yang , Haifeng Wang

Solubility of fission products (FPs) in spent nuclear fuels is a key issue in understanding their performance. In this study, a detailed density functional theory study on solubility of La, Nd, Zr and Mo in uranium dioxide (UO₂) and triuranium octoxide (U₃O₈) was carried out. The most favorable sites of La, Nd, Zr and Mo were calculated by analyzing the solution energies of these four FP elements in UO₂ and U₃O₈. Thermodynamic properties of FPs doping in UO₂ as well as in U₃O₈ were determined from 0 K to 2000 K. The results reveal that the effect of FPs on thermodynamic properties of U₃O₈ is slightly larger than that of UO₂ due to the presence of an oxygen vacancy in FP-doped U₃O₈. On this basis, the solubility of La, Nd, Zr and Mo in UO₂ and U₃O₈ was predicted, showing a good agreement with available experiments. Subsequently, the difference in solubility behaviors was studied through electronic properties. The simulation results indicate that the solubility of La, Nd, Zr and Mo in U₃O₈ is much lower than that in UO₂. Therefore, an idea to separate FPs from spent nuclear fuel is using mutual transformations between FP-doped UO₂ and FP-doped U₃O₈ under different temperatures, which is a promising way to promote the development of nuclear fuel cycles.

裂变产物（FPs）在乏核燃料中的溶解度是了解其性能的一个关键问题。本研究对 La、Nd、Zr 和 Mo 在二氧化铀（UO2）和八氧化三铀（U3O8）中的溶解度进行了详细的密度泛函理论研究。通过分析这四种元素在二氧化铀和八氧化三铀中的溶解能，计算出了 La、Nd、Zr 和 Mo 的最有利位点。结果表明，由于掺杂了 FP 的 U3O8 中存在氧空位，FP 对 U3O8 热力学性质的影响略大于 UO2。在此基础上，预测了 La、Nd、Zr 和 Mo 在二氧化铀和八氧化三铀中的溶解度，结果与现有实验吻合。随后，通过电子特性研究了溶解度行为的差异。模拟结果表明，La、Nd、Zr 和 Mo 在 U3O8 中的溶解度远远低于在二氧化钛中的溶解度。因此，利用掺杂FP的UO2和掺杂FP的U3O8在不同温度下的相互转化来分离乏核燃料中的FP是一种可行的思路，也是促进核燃料循环发展的一种可行方法。

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引用次数: 0

Effect of grain boundaries on the helium degradation mechanisms of alloy 800H: A molecular dynamics study 晶界对合金 800H 的氦降解机制的影响：分子动力学研究

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

Journal of Nuclear Materials

Pub Date : 2024-09-19 DOI: 10.1016/j.jnucmat.2024.155395

I. Cheik Njifon, E. Torres

Alloy 800H is currently used as structural material in light-water cooled nuclear reactors, and it is also considered as candidate materials for various advanced reactor designs. Under operation, the exposure of alloy 800H to neutron irradiation results in the formation of helium (He), mainly from Ni by the transmutation reaction (n, α). In this work, we present a molecular dynamics (MD) simulation study of the behavior and effects of He on the microstructure and mechanical properties of alloy 800H. Our results show that the population of clusters made of 5 to 9 He atoms is nearly constant throughout the simulation time (10 ns), while the population of larger clusters increases as the simulation time increases. The growth of clusters is controlled by either the dissociation and diffusion of smaller clusters towards nearby larger clusters or the merging of larger clusters initially located nearby to each other. A significant accumulation of He is observed at the grain boundaries (GB), while a depletion zone is found at the neighboring regions. As a result, the density of He cluster is significantly higher at the GBs as compared to the intra-granular regions. The nucleation and growth of He clusters also results in the formation of Frenkel pairs (FP), whose associated self-interstitial atoms (SIA) agglomerate into interstitial clusters in the alloy 800H matrix. As a consequence, dislocation segments, mostly of the Shockley type, are generated in the microstructure, and often located next to He clusters. The combination of the aforementioned defect structures and the high density of He clusters at the GBs results in a substantial degradation of the mechanical properties of alloy 800H single crystal and bicrystals.

合金 800H 目前用作轻水冷却核反应堆的结构材料，也被视为各种先进反应堆设计的候选材料。在运行过程中，合金 800H 受中子辐照后会形成氦（He），主要是由 Ni 通过嬗变反应（n，α）形成的。在这项工作中，我们对 He 对合金 800H 的微观结构和机械性能的行为和影响进行了分子动力学（MD）模拟研究。我们的结果表明，在整个模拟时间（10 ns）内，由 5 到 9 个 He 原子组成的团簇数量几乎是恒定的，而较大团簇的数量则随着模拟时间的延长而增加。簇群的增长受控于较小簇群的解离和向附近较大簇群的扩散，或最初位于附近的较大簇群的合并。在晶粒边界 (GB) 观察到 He 的大量积累，而在邻近区域则发现了耗竭区。因此，与晶粒内部区域相比，晶粒边界区域的氦团密度要高得多。He 簇的成核和生长还导致弗伦克尔对（FP）的形成，其相关的自间隙原子（SIA）在合金 800H 基体中聚集成间隙簇。因此，在微观结构中产生了位错段，主要是肖克利型位错段，通常位于 He 簇旁边。上述缺陷结构与 GB 上高密度的 He 簇相结合，导致合金 800H 单晶和双晶的机械性能大幅下降。

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引用次数: 0