Journal of Nuclear Materials最新文献_第5页

Point defect segregation at edge dislocations in α-Fe studied by kinetic activation-relaxation technique 用动力学激活-松弛技术研究α-Fe中边缘位错处的点缺陷偏析

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

Journal of Nuclear Materials

Pub Date : 2026-01-02 DOI: 10.1016/j.jnucmat.2026.156436

N. Kvashin , N. Anento , L. Malerba

The mechanical properties of crystalline materials such as metals, are strongly related to the mobility of dislocations, which is directly affected by their interaction with other defects present in the microstructure and acting as obstacles. Under irradiation conditions the number density of point defects increases substantially, leading to several phenomena at the atomic scale, some of which are related with the behaviour of dislocations as sinks for vacancies and self-interstitial atoms. In this work we present an in-depth study of the segregation process of point defects to an edge dislocation in α-Fe, performed with an on-the-fly kinetic Monte Carlo model, the kinetic activation-relaxation technique (k-ART). Our KMC simulations show that, in the vicinity of the dislocation core, the dynamics of vacancies and SIAs is accelerated before absorption. For the former, the preferential path is along the compression region while for the latter is along the tensile region. This work therefore provides a greater knowledge of the dynamic properties of point defects around of dislocations, such as free migration time, acceleration/deceleration of point defects motion and energies of absorption events. These results will allow more precise modelling of the microstructure evolution of polycrystalline materials, improving the predictive capabilities of existing models in the long term. In order to ensure transferability of these findings to other KMC models, the data obtained in the simulations have been used to train a prediction model based on a Machine Learning logistic regression algorithm.

金属等晶体材料的机械性能与位错的迁移率密切相关，位错的迁移率直接受其与微观结构中存在的其他缺陷的相互作用和作为障碍的影响。在辐照条件下，点缺陷的数量密度大大增加，导致原子尺度上的几种现象，其中一些与位错作为空位和自间隙原子的汇有关。在这项工作中，我们提出了一个深入的研究α-Fe中点缺陷到边缘位错的偏析过程，采用动态动力学蒙特卡罗模型，动力学激活松弛技术（k-ART）。我们的KMC模拟表明，在位错核心附近，空位和SIAs的动力学在吸收之前加速。前者的优先路径是沿压缩区，后者的优先路径是沿拉伸区。因此，这项工作为位错周围点缺陷的动态特性提供了更多的知识，例如自由迁移时间、点缺陷运动的加速/减速和吸收事件的能量。这些结果将允许对多晶材料的微观结构演变进行更精确的建模，从长远来看，提高现有模型的预测能力。为了确保这些发现可转移到其他KMC模型，在模拟中获得的数据已用于训练基于机器学习逻辑回归算法的预测模型。

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引用次数: 0

Creep constitutive model for FeCrAl alloy cladding tube: experiments and molecular dynamics simulations FeCrAl合金包层管蠕变本构模型：实验与分子动力学模拟

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

Journal of Nuclear Materials

Pub Date : 2026-01-02 DOI: 10.1016/j.jnucmat.2025.156433

Huan Yao , Changwei Wu , Tianzhou Ye , Junmei Wu , Yingwei Wu , Ping Chen , Qianjin Xia

FeCrAl alloy serves as a promising accident-tolerant fuel cladding material for nuclear reactors. The structural integrity of nuclear fuel throughout the reactor’s service life critically depends on the cladding's creep behavior. Researchers have proposed several constitutive models to predict the high-temperature creep response of FeCrAl alloy tubes. However, the development of reasonable and reliable constitutive models necessitates extensive experimental validation due to the complex interdependencies governing creep behavior. This study aims to establish a new creep constitutive model for FeCrAl alloys through molecular dynamics (MD) simulations and experimental validation. Constitutive model parameters were fitted based on the MD simulation results of high-temperature creep of FeCrAl alloy under a wide range of grain size, temperature, and stress conditions. Subsequently, the key parameters were optimized against the material's bulk mechanical properties and biaxial creep test data. The developed constitutive model was implemented in finite element analysis (FEA) to simulate biaxial creep behavior of thin-walled FeCrAl tubes. A good quantitative agreement was observed between the FEA predictions and the measured results for both axial and hoop creep strain rates. Furthermore, the model is validated against uniaxial creep data from open literature, confirming its reliability in simulating both biaxial and uniaxial creep behavior of FeCrAl cladding tubes. The proposed model can achieve at least an order of magnitude improvement in prediction accuracy over the ORNL (Oak Ridge National Laboratory) model.

铁铁合金是一种很有前途的核反应堆耐事故燃料包壳材料。在整个反应堆的使用寿命中，核燃料的结构完整性在很大程度上取决于包层的蠕变行为。研究人员提出了几种本构模型来预测FeCrAl合金管的高温蠕变响应。然而，由于控制蠕变行为的复杂相互依赖关系，开发合理可靠的本构模型需要大量的实验验证。本研究旨在通过分子动力学（MD）模拟和实验验证，建立新的FeCrAl合金蠕变本构模型。基于大范围晶粒尺寸、温度和应力条件下FeCrAl合金高温蠕变的MD模拟结果拟合本构模型参数。随后，根据材料的整体力学性能和双轴蠕变试验数据对关键参数进行了优化。将所建立的本构模型应用于薄壁FeCrAl管的双轴蠕变有限元分析。在轴向和环向蠕变应变率的有限元预测和测量结果之间观察到良好的定量一致。此外，该模型与公开文献中的单轴蠕变数据进行了验证，证实了其在模拟FeCrAl包层管双轴和单轴蠕变行为方面的可靠性。与橡树岭国家实验室（ORNL）模型相比，该模型的预测精度至少提高了一个数量级。

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引用次数: 0

Enhanced immobilization of trivalent actinides in zircon-based multiphase ceramics via spark plasma sintering 火花等离子烧结强化锆基多相陶瓷中三价锕系元素的固定化

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

Journal of Nuclear Materials

Pub Date : 2026-01-01 DOI: 10.1016/j.jnucmat.2025.156434

Yingwei Xiong , Xingtong Liu , Wenjuan Wang , Yi Ding

Ceramic immobilization is one of the good strategies for disposal of high-level radioactive waste, but suffers from high temperature and low immobilization capacity. Herein, green spark plasma sintering (SPS) technology was reported to prepare zircon-based multiphase ceramics (x-Z) for enhanced simulated trivalent actinide (Nd³⁺) immobilization. The effect of Nd³⁺ content on the phase and microstructure evolutions of the obtained x-Z ceramics was investigated. The x-Z ceramics were prepared by SPS at low sintering temperature (1350 °C) and short time (10 min), and their Nd³⁺ immobilization capacity was up to 20 at% owing to the low sintering temperature and short sintering time reduced the decomposition of ZrSiO₄ phase. In addition, the obtained x-Z ceramics exhibited superior aqueous stability due to the high density achieved by SPS. The green and efficient SPS technology could play a significant role in promoting the industrialization of ceramics immobilization of high-level radioactive waste.

陶瓷固定化是处理高放废物的一种较好的方法，但存在温度高、固定化能力低的问题。本文报道了绿色火花等离子烧结（SPS）技术制备锆基多相陶瓷（x-Z），用于增强模拟三价锕系元素（Nd3+）的固定化。研究了Nd3+含量对x-Z陶瓷相和微观结构演变的影响。采用SPS在低烧结温度（1350℃）和短烧结时间（10 min）下制备了x-Z陶瓷，由于较低的烧结温度和较短的烧结时间减少了ZrSiO4相的分解，使得x-Z陶瓷的Nd3+固定容量达到20%。此外，由于SPS获得的高密度，所获得的x-Z陶瓷具有优异的水稳定性。绿色高效的SPS技术对高放废物陶瓷固定化产业化具有重要的推动作用。

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引用次数: 0

Development of a neural network potential for osmium enables irradiation damage simulations 一个神经网络的发展潜力的锇使辐射损伤模拟

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

Journal of Nuclear Materials

Pub Date : 2025-12-30 DOI: 10.1016/j.jnucmat.2025.156424

Yu Wang , Heng Chen , Rui Su , Bin Xu , Rulong Zhou , Dongdong Li , Yu-Wei You , Pengfei Guan , Changsong Liu

This work develops a high-accuracy artificial neural network (ANN) potential for osmium (Os) to enable large-scale irradiation damage simulations in fusion materials. The potential employs spherical harmonic-Chebyshev polynomial descriptors within a Behler-Parrinello neural network architecture, trained on an extensive dataset generated via density functional theory and ab initio molecular dynamics. Comprehensive validations demonstrate excellent agreement with reference calculations and experimental data across multiple properties: lattice constants of diverse crystal structures, elastic constants for hexagonal close-packed Os, dimer interactions, and defect formation energies (vacancies, interstitials, surfaces). The ANN potential accurately reproduces key behaviors under extreme conditions, including melting characteristics, sputtering thresholds, and primary knock-on atom collision cascades. Simulations reveal defect evolution and clustering during radiation events. This transferable potential provides a critical computational tool for investigating Os precipitation effects on tritium retention and irradiation hardening in tungsten-based plasma-facing materials for fusion reactors.

这项工作为锇（Os）开发了一个高精度的人工神经网络（ANN）潜力，使聚变材料中的大规模辐射损伤模拟成为可能。该系统在Behler-Parrinello神经网络架构中使用球面谐波-切比雪夫多项式描述符，并在密度泛函理论和从头算分子动力学生成的广泛数据集上进行训练。综合验证证明了与参考计算和实验数据在多个性质上的良好一致性：不同晶体结构的晶格常数，六边形紧密排列的o的弹性常数，二聚体相互作用和缺陷形成能量（空位，间隙，表面）。人工神经网络电位精确地再现了极端条件下的关键行为，包括熔化特征、溅射阈值和初级原子碰撞级联。模拟结果揭示了辐射过程中缺陷的演化和聚类。这种可转移电位为研究聚变反应堆用钨基等离子体材料中Os沉淀对氚保留和辐照硬化的影响提供了一个重要的计算工具。

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引用次数: 0

Small-scale mechanical testing of interfacial toughness in Cr-coated zircaloy-4 cr包覆锆合金-4界面韧性的小规模力学试验

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

Journal of Nuclear Materials

Pub Date : 2025-12-30 DOI: 10.1016/j.jnucmat.2025.156426

Jiwon Mun , JungHun Park , JongDae Hong , Sebastian Lam , Peter Hosemann , Gi-dong Sim , Ho Jin Ryu

This study quantifies fracture mechanisms and interfacial fracture toughness in Cr-coated Zircaloy-4 tube systems proposed as accident-tolerant fuel (ATF) cladding, using deep-notched (DN) microtensile specimens. Three gauge configurations were tested: single-phase Zr (DN-Zr), single-phase Cr (DN-Cr), and a Zr/Cr interface located at the notch root (DN-interface). The mode I stress intensity factor (SIF)

K_{I}

for the ideal sharp-notch geometry was obtained from an analytical correlation and calibrated by finite-element (FE) J-integral analysis (Contour Integral method using ABAQUS/Standard), showing close agreement and validating the modeling. Using this calibration, the interface fracture toughness for the DN-interface configuration

K_{Q, int}

was extracted. Interfacial failure exhibits two distinct modes: specimens exhibiting interface-crossing yield

K_{Q, int}

= 1.89 ± 0.20 MPa

\sqrt m

, while delamination-dominated specimens yield

K_{Q, int}

= 1.12 ± 0.02 MPa

\sqrt m

. For safety assessments of Cr-coated Zircaloy-4 applications, we suggest a conservative, weakest-mode design input of

K_{Q, int}^{cons}

= 1.12 MPa

\sqrt m

. The calibrated methodology, combined with the conservative interfacial toughness

K_{Q, int}^{cons}

, enables quantitative screening and optimization of coating–interface configurations for ATF cladding.

本研究使用深缺口（DN）微拉伸试样，量化了作为耐事故燃料（ATF）包覆层的cr涂层锆合金-4管系统的断裂机制和界面断裂韧性。测试了三种压力表配置：单相Zr （DN-Zr）、单相Cr （DN-Cr）和位于缺口根部的Zr/Cr界面（dn -界面）。通过解析关联得到理想锐缺口几何形状的I型应力强度因子（SIF） KI，并通过有限元（FE） j积分分析（使用ABAQUS/Standard的轮廓积分法）进行校准，结果显示出非常接近的一致性，验证了模型的有效性。利用此标定，提取了dn -界面构型的界面断裂韧性KQ，int。界面破坏表现出两种不同的模式：界面交叉试样的屈服量KQ，int = 1.89±0.20 MPa√m，而分层主导试样的屈服量KQ，int = 1.12±0.02 MPa√m。对于cr涂层锆合金-4应用的安全性评估，我们建议采用保守的最弱模式设计输入KQ，intcons = 1.12 MPa / m。校准的方法，结合保守的界面韧性KQ，intcons，可以定量筛选和优化ATF包层的涂层界面配置。

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引用次数: 0

Pyrophoricity of uranium and uranium compounds: Mechanisms, knowledge gaps, and implications for nuclear safety 铀和铀化合物的焦性：机制、知识空白和对核安全的影响

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

Journal of Nuclear Materials

Pub Date : 2025-12-30 DOI: 10.1016/j.jnucmat.2025.156425

Alisha J. Cramer , Peter G. Martin , Thomas B. Scott

Depleted uranium materials pose a significant challenge with respect to their safe, long-term storage, not only from a radiological standpoint, but also with regards to fire safety owing to the reactivity of finely-divided uranium material with air. The risk of pyrophoric ignition is known to be strongly dependent on the material’s specific surface area, however, other factors are known to likely play a role in either increasing or decreasing such a risk. In this article, the research to date on the pyrophoric behaviour of uranium compounds is compiled and examined with a view to understanding the factors controlling pyrophoric ignition and determining the gaps in current knowledge. Although some influencing factors, such as specific surface area, dominate ignition behaviour, several other factors can disrupt expected behaviour, demonstrating that pyrophoric ignition is a complex, dynamic interplay of different factors.

贫铀材料不仅从放射角度，而且由于细分化铀材料与空气的反应性，对其安全、长期储存构成重大挑战。众所周知，焦燃着火的风险在很大程度上取决于材料的比表面积，然而，已知其他因素可能在增加或减少这种风险方面发挥作用。在本文中，对迄今为止关于铀化合物的热解行为的研究进行了汇编和检查，以期了解控制热解着火的因素并确定当前知识的空白。虽然一些影响因素，如比表面积，主导着点火行为，但其他一些因素可能会破坏预期的行为，这表明焦焰点火是一个复杂的，不同因素的动态相互作用。

引用次数: 0

Interfacial dislocation engineering in copper-graphene composites: Atomic insights into enhanced radiation resistance 铜-石墨烯复合材料的界面位错工程：增强抗辐射性的原子见解

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

Journal of Nuclear Materials

Pub Date : 2025-12-29 DOI: 10.1016/j.jnucmat.2025.156423

Qi Zhang, Zhuoxin Yan, Zhe Yan, Boan Zhong, Mingyu Gong, Yue Liu, Tongxiang Fan

Copper (Cu)-based materials suffer from performance degradation due to the accumulation of radiation-induced defects, limiting their application in high-radiation environments. The van der Waals (vdW) type interface formed by introducing graphene (Gr) into Cu-based materials is expected to address the challenge of modifying dislocation characteristics in metallic interface engineering. However, specific methods to modify the interface and their impacts on radiation resistance still need further investigation and quantification. Here, we investigate the interfacial vdW dislocation in Cu/Gr composites, in order to enhance radiation resistance via atomic simulations. The results reveal that increasing Gr rotation angle can lead to a rise in the interfacial dislocation density, while increasing Gr thickness broadens dislocation core width. Quantitative analysis reveals the optimal structural parameters: a critical 15° rotation angle and 4-layer thickness of Gr correspond to point defects reductions of up to 19.5 % and 35.6 %, respectively, compared to the unmodified interface. These findings are crucial for understanding and designing new vdW/metal composites with enhanced radiation resistance.

由于辐射缺陷的积累，铜基材料的性能会下降，限制了其在高辐射环境中的应用。将石墨烯（Gr）引入cu基材料形成的范德华（vdW）型界面有望解决金属界面工程中改变位错特征的挑战。然而，具体的界面修饰方法及其对抗辐射性能的影响还需要进一步的研究和量化。本文通过原子模拟研究了Cu/Gr复合材料中界面vdW位错，以提高其抗辐射能力。结果表明，增加Gr旋转角度可导致界面位错密度增大，而增加Gr厚度可使位错芯宽度变宽。定量分析表明，优化后的结构参数为临界15°旋转角度和4层Gr厚度，与未改性界面相比，点缺陷分别减少了19.5%和35.6%。这些发现对于理解和设计具有增强抗辐射性能的新型vdW/金属复合材料至关重要。

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引用次数: 0

Corrosion study of zirconium alloys via molecular dynamics simulations: I. oxygen migration in zirconium oxides under low-dose irradiation induced displacement damage 基于分子动力学模拟的锆合金腐蚀研究：1 .低剂量辐照下氧化锆中的氧迁移引起的位移损伤

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

Journal of Nuclear Materials

Pub Date : 2025-12-29 DOI: 10.1016/j.jnucmat.2025.156419

Xiang Li , Jinsong Zhang , Shuang Dai , Ke Wang , Yi Wang , Jia Tang , Shubo Yang , Qi Cao

In zirconium-alloy corrosion models, O²⁻ movement through the zirconium oxide film is commonly considered as the rate-determining step. SEM-EDS results in this study indicate that O²⁻ movement is driven by the built-in electric field rather than by concentration gradients. Molecular dynamics simulations were employed to investigate O²⁻ mobility in several zirconium oxide structures, showing that monoclinic ZrO₂ with a vertical grain boundary provides the highest migration ability. To investigate irradiation effects, the study focuses on low-dose displacement damage resulting from primary knock-on atom events. The simulations including mean square displacement, trajectory and diffusion coefficient results demonstrate that such low dose irradiation induced damage generally causes minimal change in O²⁻ mobility, suggesting a limited influence on corrosion under these conditions.

在锆合金腐蚀模型中，O2−通过氧化锆膜的运动通常被认为是速率决定步骤。SEM-EDS结果表明，O2−的运动是由内置电场驱动的，而不是由浓度梯度驱动的。分子动力学模拟研究了几种氧化锆结构中O2−迁移率，结果表明，具有垂直晶界的单斜ZrO2具有最高的迁移能力。为了研究辐照效应，研究重点是由初级原子撞击事件引起的低剂量位移损伤。包括均方位移、轨迹和扩散系数在内的模拟结果表明，这种低剂量辐射引起的损伤通常对O2−迁移率的变化很小，表明在这种条件下对腐蚀的影响有限。

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引用次数: 0

Influence of multi-ion concentrations on corrosion product deposition on zirconium alloy cladding in high-temperature water 多离子浓度对高温水中锆合金熔覆层腐蚀产物沉积的影响

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

Journal of Nuclear Materials

Pub Date : 2025-12-29 DOI: 10.1016/j.jnucmat.2025.156420

Yu Yang , Meijiao Huang , Jixue Sui , An Li , Guangming Shen , Xiaoyong Wu , Lu Wu , Mingzhang Lin

The concentrations of soluble metal ions in the primary water circulation loop of pressurized water reactors can greatly affect the deposition behaviors of corrosion products on zirconium alloy cladding tubes at high temperatures, which is referred to as CRUD. However, while considerable progress has been made in the development of CRUD deposition models, the applicability of these models remains limited, and the impacts of Fe³⁺, Ni²⁺, Cr³⁺, Co²⁺, and Mn²⁺ on the microstructure of CRUD and the corrosion mechanism have not been fully incorporated into model development. The present work addresses these issues by combining experimental characterization with thermodynamic calculations to analyze the mechanism by which high-temperature aqueous solutions with different Fe³⁺, Ni²⁺, Cr³⁺, Co²⁺, and Mn²⁺ ion concentrations influence the CRUD deposition process on zirconium alloy cladding tubes. In addition to the composition and structure of CRUD deposits, the oxide film properties of the cladding tubes are further examined under subcooled nucleate boiling (SNB) and non-SNB conditions. The results indicate that increases in the weight of zirconium alloy cladding tubes under the non-SNB condition are due primarily to oxide film thickening under low metal ion concentrations. At high ion concentrations, a thinner oxide film was observed, which is likely attributed to modifications in the local chemical environment at the metal/oxide interface induced by the CRUD deposits. In contrast, the CRUD morphology changed under the SNB condition from cluster-like deposits without chimneys to porous structures with chimneys with increasing metal ion concentrations. Experimental results obtained under the SNB condition, in addition to the results of thermodynamic calculations of the Gibbs free energy, demonstrate that Fe₃O₄ and spinel compounds, such as CoFe₂O₄ and FeCr₂O₄, are the dominant phases within the CRUD deposits.

压水堆一次水循环回路中可溶金属离子的浓度对锆合金包层管腐蚀产物在高温下的沉积行为有很大的影响，称为CRUD。然而，尽管CRUD沉积模型的开发取得了相当大的进展，但这些模型的适用性仍然有限，并且Fe3+， Ni2+, Cr3+， Co2+和Mn2+对CRUD微观结构和腐蚀机制的影响尚未完全纳入模型开发中。本研究将实验表征与热力学计算相结合，分析了不同Fe3+、Ni2+、Cr3+、Co2+和Mn2+离子浓度的高温水溶液对锆合金包层管上CRUD沉积过程的影响机制。除了CRUD沉积物的组成和结构外，还进一步研究了过冷核沸腾（SNB）和非SNB条件下包层管的氧化膜性能。结果表明，在非snb条件下，锆合金包层管重量的增加主要是由于低金属离子浓度下氧化膜增厚所致。在高离子浓度下，观察到较薄的氧化膜，这可能归因于由CRUD沉积引起的金属/氧化物界面局部化学环境的改变。而在SNB条件下，随着金属离子浓度的增加，CRUD的形貌由无烟囱的簇状沉积转变为有烟囱的多孔结构。在SNB条件下的实验结果以及Gibbs自由能的热力学计算结果表明，Fe3O4和尖晶石化合物（如CoFe2O4和FeCr2O4）是CRUD沉积层中的主要相。

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引用次数: 0

In – situ chlorination of simulated fission products from light water reactor fuel using ZrCl4 使用ZrCl4对轻水反应堆燃料模拟裂变产物进行原位氯化

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

Journal of Nuclear Materials

Pub Date : 2025-12-29 DOI: 10.1016/j.jnucmat.2025.156418

Diallo Barnes, Courtney Eckley, Mary Cernyar, Michael F. Simpson

To develop a process for chlorinating spent nuclear fuel (SNF) from light water reactors and dissolving it into a molten salt, the reactivity of several surrogate fission products (SFPs) with NiCl₂ and ZrCl₄ in molten LiCl – KCl was tested. ZrCl₄ was created in-situ by reacting NiCl₂ and Zr metal in the molten salt. Powders of SrO, La₂O₃, CeO₂, Cs₂O, and a Mo metal rod were all immersed in eutectic LiCl – KCl with initially 8.9 wt% NiCl₂, likely exceeding the solubility limit. Salt samples were collected at several time intervals over 24 h and analyzed via inductively coupled plasma mass spectrometry (ICP – MS). NiCl₂ did not directly react with any of the SFPs, but in situ formed ZrCl₄ was reactive towards all the oxides. The Mo metal rod was virtually inert towards reaction with both NiCl₂ and ZrCl₄. Experiments were run at 773 or 873 K with or without stirring condition. Conversion of SrO, CeO₂, and La₂O₃ to soluble chlorides ranged from 87 – 93 %, while Cs₂O conversion was only 64 % on average. Some of the Ni metal by-product plated onto the Mo rod, which was unreactive towards chlorination. A solid containing Ni and ZrO₂ settled to the bottom of the crucible.

为了开发一种将轻水反应堆中的乏核燃料（SNF）氯化并溶解到熔盐中的工艺，测试了几种替代裂变产物（SFPs）与NiCl2和ZrCl4在熔融LiCl - KCl中的反应性。通过NiCl2和金属Zr在熔盐中原位反应生成ZrCl4。SrO、La2O3、CeO2、Cs2O粉末和Mo金属棒均浸在共晶LiCl - KCl中，初始NiCl2含量为8.9 wt%，可能超过了溶解度极限。在24 h的时间间隔内收集盐样品，并通过电感耦合等离子体质谱（ICP - MS）进行分析。NiCl2不与任何一种SFPs直接反应，但原位形成的ZrCl4对所有氧化物都有反应。Mo金属棒对NiCl2和ZrCl4的反应几乎是惰性的。实验温度分别为773、873 K，有或无搅拌条件。SrO、CeO2和La2O3对可溶性氯化物的转化率为87% ~ 93%，而Cs2O的平均转化率仅为64%。部分镍金属副产物镀在钼棒上，对氯化反应无反应。含有Ni和ZrO2的固体沉降到坩埚的底部。

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引用次数: 0