Journal of Nuclear Materials最新文献_第9页

Study on long-term alteration behavior of simulated sulfate-bearing HLLW waste glass under thermal–hydrological–mechanical–chemical multi-field conditions 热-水文-机械-化学多场条件下模拟含硫酸盐高废玻璃长期蚀变行为研究

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

Journal of Nuclear Materials

Pub Date : 2025-12-13 DOI: 10.1016/j.jnucmat.2025.156379

Zhengdi Jiang , Xiaolin Yin , Jiaxin Huang , Ya Li , Liguo Xu , Lang Wu

Glass waste forms are at risk of groundwater intrusion during long-term geological disposal, where direct contact compromises chemical durability and may release radionuclides into the biosphere, thus necessitating a critical assessment of their chemical stability in aqueous environments. This study investigated the chemical stability of simulated sulfate-bearing high-level liquid waste (HLLW) glass under thermal–hydrological–mechanical–chemical (THMC) multi-field conditions (90°C, 0.01 mL/min flow rate, 10 MPa, in simulated groundwater) through 364-day multi-stage leaching tests. Results revealed sequential precipitation of platy BaSO₄ (7–14 days), Mg-Al-rich layered silicate (at 14 days), and acicular/prismatic CaCO₃ crystals (by 364 days). Alteration layer development initiated between 14 and 56 days (reaching 23 μm by 56 days) and thickened to 135.6 μm by 364 days, comprising three distinct zones: an innermost amorphous aluminosilicate gel layer, Mg-Al-rich silicates (containing BaSO₄), and an outermost CaCO₃ layer observed at 364 days. Dissolution rates exhibited a multi-stage evolution: rapid increase (1–3 days), decelerated increase (3–14 days), sharp decline (14–56 days), a stabilization trend (56–182 days), and the near-achievement of dissolution equilibrium (182–364 days). These findings offer important insights into the evolution of waste glass alteration under THMC multi-field conditions, yielding key safety assessment data for high-level radioactive waste disposal.

在长期地质处置过程中，玻璃废物形式有地下水侵入的危险，直接接触会损害化学耐久性，并可能向生物圈释放放射性核素，因此需要对其在水环境中的化学稳定性进行严格评估。本研究通过364天的多阶段浸出试验，研究了模拟含硫酸盐高放废液（HLLW）玻璃在热-水文-机械-化学（THMC）多场条件下（90℃，0.01 mL/min流速，10 MPa，模拟地下水中）的化学稳定性。结果显示，连续析出板状BaSO4（7-14天）、富镁铝层状硅酸盐（14天）和针状/棱柱状CaCO3晶体（364天）。蚀变层的发育始于14 ~ 56天（56天达到23 μm）， 364天增厚至135.6 μm，包括三个不同的区域：最内层的无定形铝硅酸盐凝胶层、富镁铝硅酸盐（含BaSO4）和最外层的CaCO3层，364天观察到。溶解速率呈快速增加（1 ~ 3 d）、缓慢增加（3 ~ 14 d）、急剧下降（14 ~ 56 d）、趋于稳定（56 ~ 182 d）和接近溶解平衡（182 ~ 364 d）的多阶段演变。这些发现为研究THMC多场条件下废玻璃蚀变的演变提供了重要见解，为高放射性废物处置提供了关键的安全评估数据。

{"title":"Study on long-term alteration behavior of simulated sulfate-bearing HLLW waste glass under thermal–hydrological–mechanical–chemical multi-field conditions","authors":"Zhengdi Jiang , Xiaolin Yin , Jiaxin Huang , Ya Li , Liguo Xu , Lang Wu","doi":"10.1016/j.jnucmat.2025.156379","DOIUrl":"10.1016/j.jnucmat.2025.156379","url":null,"abstract":"<div><div>Glass waste forms are at risk of groundwater intrusion during long-term geological disposal, where direct contact compromises chemical durability and may release radionuclides into the biosphere, thus necessitating a critical assessment of their chemical stability in aqueous environments. This study investigated the chemical stability of simulated sulfate-bearing high-level liquid waste (HLLW) glass under thermal–hydrological–mechanical–chemical (THMC) multi-field conditions (90°C, 0.01 mL/min flow rate, 10 MPa, in simulated groundwater) through 364-day multi-stage leaching tests. Results revealed sequential precipitation of platy BaSO<sub>4</sub> (7–14 days), Mg-Al-rich layered silicate (at 14 days), and acicular/prismatic CaCO<sub>3</sub> crystals (by 364 days). Alteration layer development initiated between 14 and 56 days (reaching 23 μm by 56 days) and thickened to 135.6 μm by 364 days, comprising three distinct zones: an innermost amorphous aluminosilicate gel layer, Mg-Al-rich silicates (containing BaSO<sub>4</sub>), and an outermost CaCO<sub>3</sub> layer observed at 364 days. Dissolution rates exhibited a multi-stage evolution: rapid increase (1–3 days), decelerated increase (3–14 days), sharp decline (14–56 days), a stabilization trend (56–182 days), and the near-achievement of dissolution equilibrium (182–364 days). These findings offer important insights into the evolution of waste glass alteration under THMC multi-field conditions, yielding key safety assessment data for high-level radioactive waste disposal.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156379"},"PeriodicalIF":3.2,"publicationDate":"2025-12-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789293","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

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 : 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。

{"title":"Quantitative assessment of compositional effects on molybdenum solubility in nuclear waste glasses","authors":"Kai Xu , Ziqiang Jia , Xiangda Meng , Yujie Liu , Jing Ma","doi":"10.1016/j.jnucmat.2025.156372","DOIUrl":"10.1016/j.jnucmat.2025.156372","url":null,"abstract":"<div><div>The limited solubility of MoO<sub>3</sub> 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 MoO<sub>3</sub> 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 MoO<sub>3</sub> solubility in borosilicate glasses and compiling a dataset of 143 crucible-scale measurements. Furthermore, an empirical model was developed to predict MoO<sub>3</sub> 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, B<sub>2</sub>O<sub>3</sub>, Li<sub>2</sub>O, ZnO, V<sub>2</sub>O<sub>5</sub>, and CaO enhance MoO<sub>3</sub> solubility, whereas Na<sub>2</sub>O and Al<sub>2</sub>O<sub>3</sub> exhibit the reverse effect.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156372"},"PeriodicalIF":3.2,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789295","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

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 : 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。

{"title":"Evaluation of Gd₂O₃-Y2O3 co-stabilized zirconia as a burnable absorber for micro HTGR applications","authors":"Longwu Kang , Anzhou Qi , Wugang Fan , Zhaoquan Zhang , Xiaochuan Jiang , Guoming Liu , Xiaojiao Wang","doi":"10.1016/j.jnucmat.2025.156375","DOIUrl":"10.1016/j.jnucmat.2025.156375","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156375"},"PeriodicalIF":3.2,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789372","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

First experimental determination of the low-temperature isothermal section of the Zr-Nb-Cr ternary system at 1073 K 首次实验测定了Zr-Nb-Cr三元体系在1073 K时的低温等温截面

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

Journal of Nuclear Materials

Pub Date : 2025-12-11 DOI: 10.1016/j.jnucmat.2025.156367

Wen-bo Wang , Kang Wang , Wenfang Li , Jun Du

This study presents the first experimental determination of the 1073 K isothermal section of the Zr-Nb-Cr ternary system, identifying five single-phase (α-Zr, β-Zr, β-Nb, BCC(Cr), C15), six two-phase, and two three-phase (α-Zr + β-Zr + C15, β-Zr + β-Nb + C15) regions. The C15 phase forms a continuous solid solution, partitioning the diagram, while β-(Zr,Nb) decomposition and β-Zr eutectoid reactions drive low-temperature phase evolution. This data supports thermodynamic databases for nuclear zirconium alloy design.

本研究首次对Zr-Nb-Cr三元体系的1073 K等温截面进行了实验测定，鉴定出5个单相区（α-Zr、β-Zr、β-Nb、BCC（Cr）、C15）、6个两相区和2个三相区（α-Zr + β-Zr + C15、β-Zr + β-Nb + C15）。C15相形成连续固溶体，分划图，而β-(Zr,Nb)分解和β-Zr共析反应驱动低温相演化。这些数据支持核锆合金设计的热力学数据库。

引用次数: 0

Physical and thermal property changes under uniform oxidation in nuclear graphite 核石墨在均匀氧化下的物理和热性能变化

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

Journal of Nuclear Materials

Pub Date : 2025-12-11 DOI: 10.1016/j.jnucmat.2025.156376

Austin C. Matthews, William E. Windes

Material properties critical to graphite core performance in High Temperature Reactor (HTR) designs were measured after uniform oxidation of fine- and medium-grain nuclear graphite grades. The core structures for gas-cooled high temperature and very high temperature advanced reactor designs are composed of large nuclear graphite block components These large core components are designed to provide neutron moderation and reflection, create a large thermal sink to assist in operational control, and form the solid core structure containing the nuclear fuel, coolant channels, and the safety critical channels for control rod insertion. During operation these graphite components are subjected to a variety of different environments including irradiation, large thermal gradients, and oxidation – either through air ingress or steam during an off-normal incident. Oxidation has been shown to be a principal degradation mechanism affecting all aspects of the nuclear graphite component functions. This study addresses the underlying physical and thermal property changes of nuclear-graphite components for oxidized mass loss ranges beyond the current recommended ASME code rule limits to ensure structural integrity and optimal performance within the graphite components (a maximum mass loss = 10 %).

对高温堆（HTR）设计中石墨芯性能至关重要的材料性能进行了均匀氧化后的测量。气冷高温和极高温先进反应堆设计的堆芯结构由大型核石墨块组件组成，这些大型堆芯组件的设计目的是提供中子的调节和反射，形成一个大型热沉以辅助运行控制，并形成包含核燃料、冷却剂通道和控制棒插入的安全临界通道的固体堆芯结构。在运行过程中，这些石墨组件经受各种不同的环境，包括辐照、大热梯度和氧化——在非正常事件中通过空气进入或蒸汽氧化。氧化已被证明是影响核石墨组分功能各方面的主要降解机制。本研究解决了核石墨组件的潜在物理和热性能变化，氧化质量损失范围超过了目前推荐的ASME规范规则限制，以确保石墨组件的结构完整性和最佳性能（最大质量损失= 10%）。

{"title":"Physical and thermal property changes under uniform oxidation in nuclear graphite","authors":"Austin C. Matthews, William E. Windes","doi":"10.1016/j.jnucmat.2025.156376","DOIUrl":"10.1016/j.jnucmat.2025.156376","url":null,"abstract":"<div><div>Material properties critical to graphite core performance in High Temperature Reactor (HTR) designs were measured after uniform oxidation of fine- and medium-grain nuclear graphite grades. The core structures for gas-cooled high temperature and very high temperature advanced reactor designs are composed of large nuclear graphite block components These large core components are designed to provide neutron moderation and reflection, create a large thermal sink to assist in operational control, and form the solid core structure containing the nuclear fuel, coolant channels, and the safety critical channels for control rod insertion. During operation these graphite components are subjected to a variety of different environments including irradiation, large thermal gradients, and oxidation – either through air ingress or steam during an off-normal incident. Oxidation has been shown to be a principal degradation mechanism affecting all aspects of the nuclear graphite component functions. This study addresses the underlying physical and thermal property changes of nuclear-graphite components for oxidized mass loss ranges beyond the current recommended ASME code rule limits to ensure structural integrity and optimal performance within the graphite components (a maximum mass loss = 10 %).</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156376"},"PeriodicalIF":3.2,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789375","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Leaching behavior of HLW glass waste form in Beishan groundwater environment 北山地下水环境中高废渣玻璃形态的浸出行为

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

Journal of Nuclear Materials

Pub Date : 2025-12-11 DOI: 10.1016/j.jnucmat.2025.156373

Feng Zhiqiang , Wang Ju , Xie Jingli , Cheng Jianfeng , Lin Jie , Xie Hua

In 2021, China's first high-level radioactive waste vitrification facility commenced operation in Guangyuan, Sichuan, while the Beishan Underground Laboratory in Gansu initiated construction. This study investigates the glass waste forms produced domestically, analyzing their surface characteristics and elemental release tendencies in both deionized water and the complex hydrogeochemical environment of Beishan groundwater, which is crucial for assessing long-term disposal safety. Research results demonstrate that glass corrosion mechanisms differ significantly between deionized water and complex Beishan groundwater. In deionized water, corrosion proceeds primarily via relatively simple ion exchange and network hydrolysis. In contrast, the complex ionic environment of Beishan groundwater triggers active interface reactions, leading to the formation of various silicate precipitation layers. These layers introduce a surface "blocking-and-release" effect, making the apparent leaching behavior more complex. The initial leaching path depends on the glass's surface condition. Alkali metals enriched on as-cast samples promote simple MgO phase formation, while the rougher surface of processed samples facilitates rapid growth of complex silicates. Despite different initial paths, surface layer evolution converges after long-term leaching. Leaching rates for matrix elements rapidly decreased to 10^–1 g·m^-2·d^-1 by day 14, then slowed to 10^–2 g·m^-2·d^-1 by day 92. The trivalent simulant La exhibited a much lower and faster-declining rate, dropping to 10^–3 g·m^-2·d^-1 by day 7 and remaining at that level thereafter, showing excellent immobilization effect on actinide elements.

2021年，中国首个高放射性废物玻璃化设施在四川广元投产，甘肃北山地下实验室开工建设。研究了国内生产的玻璃废弃物形态，分析了其在去离子水和北山地下水复杂水文地球化学环境中的表面特征和元素释放趋势，这对评估北山地下水的长期处置安全性至关重要。研究结果表明，去离子水与北山复杂地下水的玻璃腐蚀机理存在显著差异。在去离子水中，腐蚀主要通过相对简单的离子交换和网络水解进行。而北山地下水复杂的离子环境引发了活跃的界面反应，形成了各种硅酸盐沉淀层。这些层引入了表面“阻塞-释放”效应，使表面的浸出行为更加复杂。最初的浸出路径取决于玻璃的表面状况。在铸态样品中富集的碱金属促进了简单的MgO相的形成，而加工样品的粗糙表面促进了复杂硅酸盐的快速生长。虽然初始路径不同，但经过长期浸出，表层演化趋于一致。基质元素的浸出速率在第14天迅速下降至10-1 g·m-2·d-1，到第92天则放缓至10-2 g·m-2·d-1。三价模拟物La的下降速率更低，下降速度更快，在第7天降至10-3 g·m-2·d-1，此后一直保持在该水平，对锕系元素具有良好的固定效果。

{"title":"Leaching behavior of HLW glass waste form in Beishan groundwater environment","authors":"Feng Zhiqiang , Wang Ju , Xie Jingli , Cheng Jianfeng , Lin Jie , Xie Hua","doi":"10.1016/j.jnucmat.2025.156373","DOIUrl":"10.1016/j.jnucmat.2025.156373","url":null,"abstract":"<div><div>In 2021, China's first high-level radioactive waste vitrification facility commenced operation in Guangyuan, Sichuan, while the Beishan Underground Laboratory in Gansu initiated construction. This study investigates the glass waste forms produced domestically, analyzing their surface characteristics and elemental release tendencies in both deionized water and the complex hydrogeochemical environment of Beishan groundwater, which is crucial for assessing long-term disposal safety. Research results demonstrate that glass corrosion mechanisms differ significantly between deionized water and complex Beishan groundwater. In deionized water, corrosion proceeds primarily via relatively simple ion exchange and network hydrolysis. In contrast, the complex ionic environment of Beishan groundwater triggers active interface reactions, leading to the formation of various silicate precipitation layers. These layers introduce a surface \"blocking-and-release\" effect, making the apparent leaching behavior more complex. The initial leaching path depends on the glass's surface condition. Alkali metals enriched on as-cast samples promote simple MgO phase formation, while the rougher surface of processed samples facilitates rapid growth of complex silicates. Despite different initial paths, surface layer evolution converges after long-term leaching. Leaching rates for matrix elements rapidly decreased to 10<sup>–1</sup> g·m<sup>-2</sup>·d<sup>-1</sup> by day 14, then slowed to 10<sup>–2</sup> g·m<sup>-2</sup>·d<sup>-1</sup> by day 92. The trivalent simulant La exhibited a much lower and faster-declining rate, dropping to 10<sup>–3</sup> g·m<sup>-2</sup>·d<sup>-1</sup> by day 7 and remaining at that level thereafter, showing excellent immobilization effect on actinide elements.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156373"},"PeriodicalIF":3.2,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789292","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Metastable grain boundary sink behavior revealed through deep-learning image analysis 深度学习图像分析揭示亚稳态晶界汇行为

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

Journal of Nuclear Materials

Pub Date : 2025-12-10 DOI: 10.1016/j.jnucmat.2025.156369

Emily Hopkins Mang , Annie K. Barnett , Sicong He , James E. Nathaniel II , Ryan Jacobs , Dane Morgan , Michael Falk , Jaime Marian , Mitra L. Taheri

Achieving radiation tolerance in crystalline materials requires advancing our understanding of defect evolution and the corresponding grain boundary (GB) response under irradiation. One strategy for realizing more radiation tolerant materials is by tailoring GBs to behave as more efficient defect sinks, however, their non-equilibrium structural response remains insufficiently resolved. In this study, we combine deep-learning enabled object detection used on in situ transmission electron microscopy experiments and multiscale modeling efforts to examine the dynamic relationship between defect microstructure and structural evolution of the GB. To investigate the underlying mechanisms, we employed molecular dynamics simulations, which support the hypothesis that self-healing through point defect (PD) emission modulates the local GB defect concentration that would cumulate as experimentally observable defect density fluctuations. In parallel, experimental observations of time-dependent GB dislocation core evolution and network formation are corroborated by molecular dynamics and physics-based dislocation loop relaxation models suggesting that non-equilibrium GBs can transition to a PD absorption regime modulated by the elastic environment. This work aims to support the idea that GB structural properties are not static, providing unique experimental evidence that non-equilibrium GB structures evolve under extreme conditions and influence the resulting radiation-induced defect microstructure.

实现晶体材料的辐射耐受需要提高我们对辐照下缺陷演变和相应晶界（GB）响应的理解。实现更耐辐射材料的一种策略是通过剪裁gb使其表现为更有效的缺陷汇，然而，它们的非平衡结构响应仍然没有得到充分解决。在本研究中，我们将深度学习支持的目标检测用于原位透射电子显微镜实验和多尺度建模工作，以研究GB缺陷微观结构与结构演变之间的动态关系。为了研究潜在的机制，我们采用了分子动力学模拟，支持了通过点缺陷（PD）发射的自我修复调节局部GB缺陷浓度的假设，该缺陷浓度会随着实验观察到的缺陷密度波动而累积。同时，分子动力学和基于物理的位错环弛豫模型证实了随时间变化的GB位错核演化和网络形成的实验观察结果，表明非平衡GB可以过渡到弹性环境调制的PD吸收状态。这项工作旨在支持GB结构性能不是静态的观点，提供了独特的实验证据，证明非平衡GB结构在极端条件下演变并影响由此产生的辐射诱导缺陷微观结构。

{"title":"Metastable grain boundary sink behavior revealed through deep-learning image analysis","authors":"Emily Hopkins Mang , Annie K. Barnett , Sicong He , James E. Nathaniel II , Ryan Jacobs , Dane Morgan , Michael Falk , Jaime Marian , Mitra L. Taheri","doi":"10.1016/j.jnucmat.2025.156369","DOIUrl":"10.1016/j.jnucmat.2025.156369","url":null,"abstract":"<div><div>Achieving radiation tolerance in crystalline materials requires advancing our understanding of defect evolution and the corresponding grain boundary (GB) response under irradiation. One strategy for realizing more radiation tolerant materials is by tailoring GBs to behave as more efficient defect sinks, however, their non-equilibrium structural response remains insufficiently resolved. In this study, we combine deep-learning enabled object detection used on <em>in situ</em> transmission electron microscopy experiments and multiscale modeling efforts to examine the dynamic relationship between defect microstructure and structural evolution of the GB. To investigate the underlying mechanisms, we employed molecular dynamics simulations, which support the hypothesis that self-healing through point defect (PD) emission modulates the local GB defect concentration that would cumulate as experimentally observable defect density fluctuations. In parallel, experimental observations of time-dependent GB dislocation core evolution and network formation are corroborated by molecular dynamics and physics-based dislocation loop relaxation models suggesting that non-equilibrium GBs can transition to a PD absorption regime modulated by the elastic environment. This work aims to support the idea that GB structural properties are not static, providing unique experimental evidence that non-equilibrium GB structures evolve under extreme conditions and influence the resulting radiation-induced defect microstructure.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156369"},"PeriodicalIF":3.2,"publicationDate":"2025-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882271","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Unveiling the origin of different fluorine-induced segregation properties of Cu and Cr on Ni-based alloy surfaces: Insights from DFT study 揭示镍基合金表面Cu和Cr不同氟致偏析性质的起源：来自DFT研究的见解

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

Journal of Nuclear Materials

Pub Date : 2025-12-10 DOI: 10.1016/j.jnucmat.2025.156370

Juntao Huang , Chunyan Yu , Jingting Luo , Yongwen Guo , Likai Guo , Jingchun Li , Yong Liu

Adsorbate-induced surface segregation is a critical behavior governing the alloy performance. In corrosive fluorine-rich environments, however, this phenomenon remains unclear. Herein, first-principles density functional theory (DFT) calculations were employed to investigate the fluorine-induced segregation of two representative alloying elements (Cu and Cr) on Ni-based alloy surfaces. An opposite effect was found that F adsorption can suppress Cu segregation while enhancing Cr segregation. Structural analysis revealed that lattice distortion alone is insufficient to account for the observed difference in segregation trends. Instead, surface electronic properties play a more dominant role. The Cr-F interaction features as strong orbital hybridization and localized charge transfer, favoring Cr segregation to the surface. Conversely, Cu shows weaker bonding with F, with partial electron transferred into adjacent Ni atoms. This indirectly results in enhanced Cu-Ni bonding along vertical direction and reduced surface stability, driving Cu to migrate into the subsurface layer. These findings unveil the atomic-level mechanisms of element-specific segregation behaviors under fluorine adsorption, and provide insights into the early-stage dealloying and corrosion processes of Ni-based alloys in fluorine-rich environments.

吸附物引起的表面偏析是控制合金性能的关键行为。然而，在腐蚀性富氟环境中，这种现象尚不清楚。本文采用第一性原理密度泛函理论（DFT）计算研究了两种具有代表性的合金元素（Cu和Cr）在ni基合金表面的氟致偏析。吸附F可以抑制Cu偏析，同时增强Cr偏析。结构分析表明，单靠晶格畸变不足以解释所观察到的偏析趋势差异。相反，表面电子性质起着更重要的作用。Cr- f相互作用表现为强轨道杂化和局域电荷转移，有利于Cr向表面偏析。相反，Cu与F的成键较弱，部分电子转移到相邻的Ni原子中。这间接导致Cu- ni沿垂直方向键合增强，表面稳定性降低，促使Cu向亚表层迁移。这些发现揭示了氟吸附下元素特异性偏析行为的原子水平机制，并为富氟环境下ni基合金的早期脱合金化和腐蚀过程提供了见解。

{"title":"Unveiling the origin of different fluorine-induced segregation properties of Cu and Cr on Ni-based alloy surfaces: Insights from DFT study","authors":"Juntao Huang , Chunyan Yu , Jingting Luo , Yongwen Guo , Likai Guo , Jingchun Li , Yong Liu","doi":"10.1016/j.jnucmat.2025.156370","DOIUrl":"10.1016/j.jnucmat.2025.156370","url":null,"abstract":"<div><div>Adsorbate-induced surface segregation is a critical behavior governing the alloy performance. In corrosive fluorine-rich environments, however, this phenomenon remains unclear. Herein, first-principles density functional theory (DFT) calculations were employed to investigate the fluorine-induced segregation of two representative alloying elements (Cu and Cr) on Ni-based alloy surfaces. An opposite effect was found that F adsorption can suppress Cu segregation while enhancing Cr segregation. Structural analysis revealed that lattice distortion alone is insufficient to account for the observed difference in segregation trends. Instead, surface electronic properties play a more dominant role. The Cr-F interaction features as strong orbital hybridization and localized charge transfer, favoring Cr segregation to the surface. Conversely, Cu shows weaker bonding with F, with partial electron transferred into adjacent Ni atoms. This indirectly results in enhanced Cu-Ni bonding along vertical direction and reduced surface stability, driving Cu to migrate into the subsurface layer. These findings unveil the atomic-level mechanisms of element-specific segregation behaviors under fluorine adsorption, and provide insights into the early-stage dealloying and corrosion processes of Ni-based alloys in fluorine-rich environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156370"},"PeriodicalIF":3.2,"publicationDate":"2025-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145786933","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Immobilization of high-sodium and cesium-rich waste derived from TRPO process in single phase hollandite ceramic waste forms TRPO工艺产生的高钠和富铯废物在单相荷兰石陶瓷废物形态中的固定化

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

Journal of Nuclear Materials

Pub Date : 2025-12-10 DOI: 10.1016/j.jnucmat.2025.156368

Zhiguo Yang , Guoqiang Zhao , Wanjun Shi , Xianzhou Ning , Bo Xie , Wei Zhang , Bin Ye , Yushan Yang

A single phase hollandite waste form was developed to immobilize the high-sodium and cesium-rich waste (HSCRW) stream derived from the trialkyl phosphine oxide (TRPO) process. In this work, the (1-x)Ba_1.2Cr_2.4Ti_5.6O₁₆·xHSCRW (0.0 ≤ x ≤ 0.2) ceramics were fabricated to investigate the effect of HSCRW incorporation on phase composition, microstructure and chemical durability of the synthesized hollandite ceramics. It was found that all waste elements are successfully embedded into the hollandite crystal structure, and the samples sintered at 1150 °C with x ≤ 0.15 showed a pure hollandite phase. The leaching test indicated that the normalized leaching rates of the waste elements Cs, Na, Rb, Sr, Mo, Fe, Ni, Ru and Rh in the as-prepared ceramic waste forms were ∼ 10^-3 g·m^-2·d^-1, with corresponding LX values > 14.5 after 28 days of leaching. Moreover, the leached samples maintained a single-phase hollandite with tetragonal structure (I4/m). These results demonstrate that hollandite ceramics can serve as promising host matrices for immobilizing HSCRW.

为固定化氧化三烷基膦（TRPO）工艺产生的高钠富铯废物（HSCRW）流，研制了一种单相荷兰酸盐废物形式。本文制备了（1-x）Ba1.2Cr2.4Ti5.6O16·xHSCRW（0.0≤x≤0.2）陶瓷，研究了HSCRW掺入对合成的荷兰石陶瓷的相组成、微观结构和化学耐久性的影响。发现所有废元素都成功嵌入到荷兰石晶体结构中，在1150℃下烧结，x≤0.15的样品显示出纯净的荷兰石相。浸出试验表明，经28 d浸出后，制备的陶瓷废渣中废元素Cs、Na、Rb、Sr、Mo、Fe、Ni、Ru、Rh的归一化浸出率为~ 10 ~ 3 g·m-2·d-1， LX值为>； 14.5。此外，浸出样品保持了单相的四边形结构（I4/m）。这些结果表明，荷兰石陶瓷可以作为固定化HSCRW的宿主基质。

{"title":"Immobilization of high-sodium and cesium-rich waste derived from TRPO process in single phase hollandite ceramic waste forms","authors":"Zhiguo Yang , Guoqiang Zhao , Wanjun Shi , Xianzhou Ning , Bo Xie , Wei Zhang , Bin Ye , Yushan Yang","doi":"10.1016/j.jnucmat.2025.156368","DOIUrl":"10.1016/j.jnucmat.2025.156368","url":null,"abstract":"<div><div>A single phase hollandite waste form was developed to immobilize the high-sodium and cesium-rich waste (HSCRW) stream derived from the trialkyl phosphine oxide (TRPO) process. In this work, the (1-<em>x</em>)Ba<sub>1.2</sub>Cr<sub>2.4</sub>Ti<sub>5.6</sub>O<sub>16</sub>·<em>x</em>HSCRW (0.0 ≤ <em>x</em> ≤ 0.2) ceramics were fabricated to investigate the effect of HSCRW incorporation on phase composition, microstructure and chemical durability of the synthesized hollandite ceramics. It was found that all waste elements are successfully embedded into the hollandite crystal structure, and the samples sintered at 1150 °C with <em>x</em> ≤ 0.15 showed a pure hollandite phase. The leaching test indicated that the normalized leaching rates of the waste elements Cs, Na, Rb, Sr, Mo, Fe, Ni, Ru and Rh in the as-prepared ceramic waste forms were ∼ 10<sup>-3</sup> <em>g</em>·m<sup>-2</sup>·d<sup>-1</sup>, with corresponding <em>LX</em> values > 14.5 after 28 days of leaching. Moreover, the leached samples maintained a single-phase hollandite with tetragonal structure (<em>I</em>4/m). These results demonstrate that hollandite ceramics can serve as promising host matrices for immobilizing HSCRW.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156368"},"PeriodicalIF":3.2,"publicationDate":"2025-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789374","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

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。

{"title":"Atomistic insights into the interfacial stability of graphene-reinforced Ni-based alloy composites after cumulative recoil events","authors":"Hai Huang , Xu Yu , Yanxin Jiang , Qing Peng , Guanyu Liu , Xiaobin Tang","doi":"10.1016/j.jnucmat.2025.156371","DOIUrl":"10.1016/j.jnucmat.2025.156371","url":null,"abstract":"<div><div>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 <em>sp<sup>3</sup></em> 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.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156371"},"PeriodicalIF":3.2,"publicationDate":"2025-12-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145786474","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0