Journal of Nuclear Materials最新文献_第8页

Synergistic roles of electronic and nuclear energy deposition: From defect generation to performance degradation in heavy-ion-irradiated CdZnTe crystals 电子和核能沉积的协同作用：从缺陷的产生到重离子辐照CdZnTe晶体的性能退化

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-19 DOI: 10.1016/j.jnucmat.2025.156393

Lu Liang , Lingyan Xu , Qinzeng Hu , Yingming Wang , Zhentao Qin , Yanyan Lei , Wei Zheng , Shuai Song , Chaopeng Mi , Roman Lanovsky , Wanqi Jie

Understanding the irradiation damage of cadmium zinc telluride (CdZnTe, CZT) crystals and its effect on photoelectric properties is crucial for their reliable use in radiation detection. This study examines the combined influence of electronic (S_e) and nuclear (S_n) energy loss on the microstructure, current transport, and carrier characteristics of CZT using 516 MeV and 1.5 MeV Xe ion irradiations. Results indicate that the synergistic S_e/S_n effect critically influences defect evolution pathways, leading to divergent microstructures. High S_e favors dislocation loops through thermal-spike-enhanced kinetics, whereas high S_n, by exacerbating lattice damage, promotes stacking faults and the evolution of loops into large-scale dislocation lines. Defect levels are deeper and the concentration of defects is larger after 516 MeV irradiation than after 1.5 MeV. Leakage current mechanisms are dominated by Schottky emission (SE) combined with the Poole-Frenkel (PF) effect for 516 MeV Xe ions, and by Fowler-Nordheim (F-N) tunneling coupled with PF effect for 1.5 MeV Xe ions. Carrier transport and γ-ray detection performance degrade more severely under 516 MeV irradiation, likely because of its broader damage layer and deeper defect levels. These findings provide a theoretical basis for understanding radiation damage mechanisms and performance recovery in CZT, offering valuable insights for radiation protection in spaceborne equipment.

了解碲化镉锌（CdZnTe， CZT）晶体的辐照损伤及其对光电性能的影响对其在辐射检测中的可靠应用至关重要。本研究利用516 MeV和1.5 MeV的Xe离子辐照，考察了电子（Se）和核（Sn）能量损失对CZT微结构、电流输运和载流子特性的综合影响。结果表明，Se/Sn的协同效应对缺陷的演化路径有重要影响，导致微观结构的分化。高Se通过热峰强化动力学有利于位错环的形成，而高Sn通过加剧晶格损伤促进层错和位错环向大规模位错线的演化。516 MeV辐照比1.5 MeV辐照后缺陷层次更深，缺陷浓度更大。泄漏电流机制主要为516 MeV Xe离子的肖特基发射（SE）和pole - frenkel （PF）效应，以及1.5 MeV Xe离子的Fowler-Nordheim （F-N）隧穿和PF效应。在516 MeV辐照下，载流子输运和γ射线探测性能下降更为严重，这可能是因为其损伤层更广，缺陷层次更深。这些发现为理解辐射损伤机理和性能恢复提供了理论基础，为星载设备的辐射防护提供了有价值的见解。

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引用次数: 0

Optimising glass-ceramic compositions for zirconolite-based actinide immobilisation 锆英石基锕系物固定化玻璃陶瓷组合物的优化

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-20 DOI: 10.1016/j.jnucmat.2025.156399

Joel L. Abraham , Pranesh Dayal , Rifat Farzana , Ghazaleh Bahmanrokh , Charles C. Sorrell , Pramod Koshy , Daniel J. Gregg

Zirconolite is a candidate wasteform for actinide immobilisation. The addition of glass to form a glass-ceramic (GC) is also under consideration as GC materials provide flexibility to immobilise heterogeneous actinide wastes and simplify processing requirements. However, a major challenge in the design of zirconolite GCs is control of the phase assemblage to minimise unwanted phase formation, particularly at high glass contents where zirconolite can be destabilised in the glass melt during consolidation. In the current research, an optimal glass composition was developed to minimise unwanted secondary phases. Initially, GCs targeting zirconolite (CaZrTi₂O₇) with varying amounts (0–100 wt%) of glass addition (NaAl_0.5B_0.5Si₂O₆) were fabricated using a pre-synthesis route. X-ray diffraction (XRD) analysis of these baseline formulations showed that undesired phases (e.g., zircon) became more apparent at higher glass contents (e.g., 75 wt%). Following this, the additions of Al₂O₃, CaO, and TiO₂ to the glass composition minimised unwanted phase formation in the GCs, including those formulations with high glass contents. The optimal glass composition was determined to be NaAl_1.5Ca_0.7Ti_0.2B_0.5Si₂O_8.6. Ce-bearing zirconolite GCs (Ca_0.8Ce_0.2ZrTi_1.6Al_0.4O₇; Ce as actinide surrogate) with varying amounts (0–100 vol%) of the tailored glass design (NaAl_1.5Ca_0.7Ti_0.2B_0.5Si₂O_8.6) were then fabricated using an in-situ crystallisation route. X-ray diffraction (XRD) and scanning electron microscopy (SEM) analyses showed that near phase-pure microstructures were achieved across all glass contents. Furthermore, the addition of glass lowered the sintering temperature (1320 °C to 1270 °C) needed to immobilise CeO₂ in zirconolite.

锆石是锕系元素固定化的候选废物。添加玻璃形成玻璃陶瓷（GC）也在考虑之中，因为GC材料提供了固定非均质锕系元素废物和简化处理要求的灵活性。然而，设计锆石gc的一个主要挑战是控制相组合，以尽量减少不必要的相形成，特别是在高玻璃含量的情况下，锆石在玻璃熔体固结过程中可能会不稳定。在目前的研究中，开发了一种最佳的玻璃成分，以尽量减少不必要的二次相。首先，采用预合成路线制备了不同玻璃添加量（NaAl0.5B0.5Si2O6）（0-100 wt%）的锆石（CaZrTi2O7）靶向gc。这些基准配方的x射线衍射（XRD）分析表明，当玻璃含量较高（例如75% wt%）时，不需要的相（例如锆石）变得更加明显。在此之后，在玻璃成分中添加Al2O3、CaO和TiO2可以最大限度地减少gc中不必要的相形成，包括那些玻璃含量高的配方。确定最佳玻璃组分为NaAl1.5Ca0.7Ti0.2B0.5Si2O8.6。采用原位晶化方法制备了含Ce锆石gc （Ca0.8Ce0.2ZrTi1.6Al0.4O7； Ce作为锕系元素替代物）和不同含量（0-100 vol%）的定制玻璃（NaAl1.5Ca0.7Ti0.2B0.5Si2O8.6）。x射线衍射（XRD）和扫描电镜（SEM）分析表明，在所有玻璃含量中都实现了接近相纯的微观结构。此外，玻璃的加入降低了在锆石中固定CeO2所需的烧结温度（1320℃至1270℃）。

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引用次数: 0

Quantitative assessment of compositional effects on molybdenum solubility in nuclear waste glasses 组分对核废料玻璃中钼溶解度影响的定量评价

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-11 DOI: 10.1016/j.jnucmat.2025.156372

Kai Xu , Ziqiang Jia , Xiangda Meng , Yujie Liu , Jing Ma

The limited solubility of MoO₃ in conventional borosilicate waste glasses can promote the formation of a molybdate-rich molten salt phase, which compromises both waste form durability and melter integrity. To avoid the accumulation of separated phases during vitrification of high-level liquid waste (HLLW), it is crucial to develop glass matrices with enhanced MoO₃ solubility. However, such development remains a challenge due to the absence of quantitative methods for evaluating the tolerance of glass compositions to molybdenum. In this study, this issue is addressed by proposing a method to quantify MoO₃ solubility in borosilicate glasses and compiling a dataset of 143 crucible-scale measurements. Furthermore, an empirical model was developed to predict MoO₃ solubility as a function of glass composition, and independent validation confirms its applicability to HLLW glasses. Although further data can improve accuracy, this model provides quantitative insights into compositional effects. The model-predicted effects of key components align with general trends previously reported in the literature. Notably, B₂O₃, Li₂O, ZnO, V₂O₅, and CaO enhance MoO₃ solubility, whereas Na₂O and Al₂O₃ exhibit the reverse effect.

MoO3在常规硼硅酸盐废玻璃中的有限溶解度会促进富钼酸盐熔融盐相的形成，从而影响废玻璃的耐久性和熔体的完整性。为了避免高放废液（HLLW）玻璃化过程中分离相的积累，开发具有增强MoO3溶解度的玻璃基质至关重要。然而，这种发展仍然是一个挑战，因为缺乏定量的方法来评估玻璃组合物对钼的耐受性。本研究提出了一种量化硼硅酸盐玻璃中MoO3溶解度的方法，并编制了143个坩埚尺度测量数据集，解决了这一问题。此外，建立了一个经验模型来预测MoO3溶解度与玻璃成分的关系，并进行了独立验证，证实了该模型适用于HLLW玻璃。虽然进一步的数据可以提高准确性，但该模型提供了对构图效果的定量见解。模型预测的关键成分的影响与文献中先前报道的一般趋势一致。值得注意的是，B2O3、Li2O、ZnO、V2O5和CaO提高了MoO3的溶解度，而Na2O和Al2O3则相反。

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引用次数: 0

Research on the regulation mechanism of laser shock peening on SCC performance of thin-walled austenitic welded joints 激光冲击强化对薄壁奥氏体焊接接头SCC性能的调节机理研究

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-13 DOI: 10.1016/j.jnucmat.2025.156381

Qianwu Li, Chengtao Li, Jing Wan, Zhaoguang Zhu, Zhilin Chen, Shugang Zhang, Bin Yang, Yuanbin Gui

To address the susceptibility of thin-walled austenitic stainless steel welded joints in nuclear power plants to stress corrosion cracking (SCC) under multi-field coupling environments, this study employs laser shock peening (LSP) technology for surface treatment to enhance their SCC resistance. By controlling LSP parameters, the research achieved redistribution of residual stresses and the construction of gradient nanostructures, effectively suppressing the initiation and propagation of SCC. Experimental results indicate that LSP treatment transforms residual tensile stresses on the welded joint surface into compressive stresses, significantly reducing the driving force for crack propagation. Simultaneously, the formed gradient nanostructure further mitigates stress concentration, thereby enhancing the material’s crack resistance. EDS analysis reveals that LSP treatment induces redistribution of Cr and Ni elements at grain boundaries, increasing precipitation of M₂₃C₆-type carbides. Although local chromium-depleted zones may elevate the risk of microcrack initiation, the strengthening effect of the gradient nanostructure predominantly governs the overall improvement in SCC resistance. Furthermore, optimized LSP parameters for thin-walled structures were established, and a three-stage strengthening model was developed, providing theoretical guidance for selecting optimal process parameters in practical applications. This research offers new technological pathways and theoretical support for enhancing SCC protection in critical thin-walled welded components of nuclear power plants.

针对核电站薄壁奥氏体不锈钢焊接接头在多场耦合环境下易发生应力腐蚀开裂（SCC）的问题，采用激光冲击强化（LSP）技术进行表面处理，提高其抗应力腐蚀开裂能力。通过对LSP参数的控制，实现了残余应力的重新分布和梯度纳米结构的构建，有效抑制了SCC的产生和扩展。实验结果表明，LSP处理将焊接接头表面的残余拉应力转化为压应力，显著降低了裂纹扩展的驱动力。同时，形成的梯度纳米结构进一步减轻了应力集中，从而提高了材料的抗裂性。EDS分析表明，LSP处理导致Cr和Ni元素在晶界处重新分布，增加了m23c6型碳化物的析出。虽然局部缺铬区可能会增加微裂纹萌生的风险，但梯度纳米结构的强化效应主导了抗SCC性能的整体提高。建立了薄壁结构的优化LSP参数，建立了三阶段强化模型，为实际应用中选择最优工艺参数提供了理论指导。本研究为加强核电站关键薄壁焊接构件的SCC保护提供了新的技术途径和理论支持。

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引用次数: 0

Effect of D on the microstructure evolution and hardness of W-Re films deposited by magnetron sputtering D对磁控溅射W-Re薄膜显微组织演变及硬度的影响

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-19 DOI: 10.1016/j.jnucmat.2025.156395

Wenjie Zhang, Zhenyu Jiang, Honghui Zhang, Tianyi Song, Yong Liu, Yao Zhang, Ze Li, Yubin Pan, Kaigui Zhu

One of the challenges in fusion reactor research is the development of plasma-facing materials that can endure extreme environments. Although tungsten (W) is regarded as the most promising plasma-facing material, its performance inevitably degrades under fusion reactor conditions. When exposed to low-energy, high-flux hydrogen isotope plasma, W will experience a risk of structural damage. Meanwhile, neutron irradiation will generate transmutation products of W, which can alter the mechanical performance of W alloys. In this work, W-rhenium (Re) films with homogeneous deuterium (D) distribution are manufactured by magnetron sputtering to investigate the effects of Re and D on the mechanical performance of W. The damage gradient effect of the prepared samples is effectively mitigated, which has significant advantages compared to samples prepared by ion implantation or plasma exposure. The results indicate that sample hardness would increase with the higher D content, and the hardness-depth relationship follows the Nix-Gao model. At the same D concentration, increasing Re content leads to a reduction in hardness, which can be attributed to enhanced dislocation mobility. Meanwhile, the dislocation density in the sample has a positive relation with D content, which is consistent with the calculation results based on dispersed barrier-hardening model. The increased Re content further reduces D retention in W-Re films by decreasing the concentration of vacancy defects. This study elucidates the effects of Re and D on the mechanical properties of W under fusion reactor conditions.

聚变反应堆研究的挑战之一是开发能够承受极端环境的面向等离子体的材料。虽然钨(W)被认为是最有前途的等离子体材料，但在聚变反应堆条件下，其性能不可避免地会下降。当暴露在低能量、高通量的氢同位素等离子体中时，W将面临结构损伤的风险。同时，中子辐照会产生W的嬗变产物，改变W合金的力学性能。本文采用磁控溅射法制备了氘(D)分布均匀的w -铼（Re）薄膜，研究了Re和D对w力学性能的影响。制备的样品有效地减轻了损伤梯度效应，与离子注入或等离子体暴露制备的样品相比具有显著的优势。结果表明，样品硬度随D含量的增加而增加，硬度与深度的关系符合Nix-Gao模型。在相同的D浓度下，稀土含量的增加导致硬度降低，这可归因于位错迁移率的增强。同时，样品中的位错密度与D含量呈正相关，这与基于分散势垒硬化模型的计算结果一致。Re含量的增加通过降低空位缺陷的浓度进一步降低了W-Re薄膜中的D保留。本研究阐明了Re和D对W在核聚变条件下力学性能的影响。

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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 : 2026-02-01 Epub 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 : 2026-02-01 Epub 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

Evaluation of Gd₂O₃-Y2O3 co-stabilized zirconia as a burnable absorber for micro HTGR applications Gd₂O₃-Y2O3共稳定氧化锆作为微HTGR可燃吸收剂的评价

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-11 DOI: 10.1016/j.jnucmat.2025.156375

Longwu Kang , Anzhou Qi , Wugang Fan , Zhaoquan Zhang , Xiaochuan Jiang , Guoming Liu , Xiaojiao Wang

Gadolinium is a well-known neutron-absorbing nuclide, yet its optimal form as a burnable absorber for micro HTGRs (high-temperature gas-cooled reactors) has not been determined. In this work, a Gd₂O₃-Y₂O₃ co-stabilized zirconia (GdY-FSZ) is explored as a promising burnable absorber by systematically investigating the temperature-dependent properties relevant to micro HTGR applications. Reactivity simulation using the Monte Carlo code RMC demonstrates that Gd₂O₃ effectively controls excess reactivity without reactivity penalty at end-of-life. The sintered GdY-FSZ exhibits a stable cubic phase structure and develops a grayish discoloration after annealing under simulated core conditions. At 1273 K, GdY-FSZ demonstrates an elastic modulus of 153 GPa and a compressive strength of 455 MPa, exceeding the ASTM C1066 specification for nuclear-grade ZrO₂ pellets. Oxygen vacancy activation near 873 K significantly influences temperature-dependent variations in elastic modulus and may also affect thermal conductivity. The latter varies from 2.5 W/(m·K) to 1.99 W/(m·K) from room temperature (RT) to 1273 K. The thermal expansion coefficients increase from 8.51 to 10.82 × 10⁻⁶ K⁻¹, eliminating the risk of mechanical interference with the graphite channels. The TG-DSC curve of GdY-FSZ demonstrates phase stability up to 1273 K, with heat flow trends associated with its thermophysical properties. Thermal shock resistance tests show a 25 % residual strength retention after two cycles from 1273 K to RT, remaining structurally stable under operational temperature fluctuations (e.g., reactor startup/shutdown). Infrared emissivity analysis across 3.3–25 μm indicates decreasing average emissivity with temperature, thereby providing essential data for heat transfer simulations preceding neutron irradiation tests. These data support the application of GdY-FSZ in a micro HTGR with graphite core and offer theoretical guidelines for new burnable absorber design.

钆是一种众所周知的中子吸收核素，但它作为微型高温气冷堆（htgr）可燃吸收剂的最佳形式尚未确定。在这项工作中，通过系统地研究与微HTGR应用相关的温度依赖特性，探索了Gd₂O₃-Y₂O₃共稳定氧化锆（GdY-FSZ）作为一种有前途的可燃吸收剂。使用蒙特卡罗代码RMC的反应性模拟表明，Gd₂O₃有效地控制了过量的反应性，而在生命周期结束时没有反应性损失。烧结后的GdY-FSZ具有稳定的立方相结构，在模拟堆芯条件下退火后呈现灰色变色。在1273 K时，GdY-FSZ的弹性模量为153 GPa，抗压强度为455 MPa，超过了ASTM C1066对核级ZrO₂球团的规范。873 K附近的氧空位活化显著影响弹性模量的温度依赖性变化，也可能影响热导率。从室温（RT）到1273 K，后者的变化范围为2.5 W/(m·K) ~ 1.99 W/（m·K）。热膨胀系数从8.51增加到10.82 × 10⁻⁶K⁻¹，消除了石墨通道受到机械干扰的风险。GdY-FSZ的TG-DSC曲线表明，GdY-FSZ的相稳定性高达1273 K，热流趋势与其热物性相关。耐热冲击测试表明，从1273 K到RT两个循环后，残余强度保持25%，在操作温度波动（例如反应堆启动/关闭）下保持结构稳定。在3.3 ~ 25 μm范围内的红外发射率分析表明，平均发射率随温度的升高而降低，为中子辐照试验前的传热模拟提供了必要的数据。这些数据支持GdY-FSZ在石墨芯微型高温堆中的应用，并为新型可燃吸收器的设计提供理论指导。

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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 : 2026-02-01 Epub 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-resolution characterization of ceramic-metal interface of TiN coating on ferritic-steels for nuclear application 核用铁素体钢TiN涂层陶瓷-金属界面的高分辨率表征

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-23 DOI: 10.1016/j.jnucmat.2025.156404

Yuhan Li , Chao Jiang , Tiankai Yao , Haiyan Wang , Jian Gan

Advanced fuel cladding is critical for fast reactors, offering sufficient thermal conductivity, mechanical and dimensional stability and radiation tolerance of the cladding base material. Additionally, it must provide corrosion resistance and high temperature coolant compatibility on the cladding outer surface, as well as chemical stability on the cladding inner wall against fuel cladding chemical interaction (FCCI). TiN ceramic coating has been considered an effective diffusion barrier for inner and outer cladding-walls for enhanced performance. The TiN-metal interface microstructure and chemistry play a critical role in coating bond strength and integrity under harsh conditions. High-resolution transmission electron microscopy characterization of ceramic-metal interface at atomic resolution in unirradiated, irradiated and thermal cycled conditions were performed. The interface remained intact after irradiation up to 200 dpa or thermal cycling five times up to 550 °C. This work discusses the potential impact of these results on coating performance and design for advanced claddings.

先进的燃料包壳对快堆至关重要，它提供了足够的导热性、机械和尺寸稳定性以及包壳基材的辐射容忍度。此外，它必须在包层外表面提供耐腐蚀性和高温冷却剂兼容性，以及在包层内壁上防止燃料包层化学相互作用（FCCI）的化学稳定性。TiN陶瓷涂层被认为是一种有效的内外包层扩散屏障，可以提高性能。在恶劣条件下，tin -金属界面的微观结构和化学性质对镀层的结合强度和完整性起着至关重要的作用。采用高分辨率透射电镜对未辐照、辐照和热循环条件下的陶瓷-金属界面进行了原子分辨率表征。在高达200 dpa的辐照或高达550°C的热循环5次后，界面保持完整。这项工作讨论了这些结果对涂层性能和先进包层设计的潜在影响。

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