Journal of Nuclear Materials最新文献_第10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 : 2026-02-01 Epub 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。

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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 : 2026-02-01 Epub 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。

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引用次数: 0

Evaluation of damage effects in 304 high-boron austenitic stainless steel under only and simultaneous Fe, He ions irradiation 304高硼奥氏体不锈钢在单一和同时铁、氦离子照射下的损伤效果评价

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-14 DOI: 10.1016/j.jnucmat.2025.156378

Jinhua Hao , Peng Zhang , Xi Li , Bin Wang , Zhihui Cai , Lifeng Ma , Xianxiu Mei

The introduction of precipitates into austenitic stainless steel is an effective strategy for improving irradiation resistance. This study employed only and simultaneous Fe, He ions irradiation on 304 high-boron austenitic stainless steel at room temperature and investigated the mechanism of ion irradiation on irradiation-induced defects and mechanical property. Throughout the ion penetration region under different irradiation conditions, both the austenite and TiB₂ phase retained original crystal structure, while the (Fe,Cr)₂B phase underwent a crystal-to-amorphous transformation. Only He ions irradiation generated nanoscale helium bubbles and induced a high density of fine dislocation loops in the damage peak region within the austenitic matrix. Under simultaneous Fe+He ions irradiation, the average size of helium bubbles and the width of helium bubbles layer increased, and at the same time, the incorporation of Fe ions also promoted the coalescence and growth of dislocation loops. The amorphization of the (Fe,Cr)₂B phase after ion irradiation played a dominant role in softening, while the pinning effect of helium bubbles inhibited hardness decrease. Irradiation-induced dislocation loops played a major role in hardening, especially the small and dense dislocation loops induced by only He ions irradiation generated significant hardening effects. Under simultaneous Fe+He ions irradiation, the 304 high-boron stainless steel with diffusely distributed borides exhibited slight softening without evident hardening. The behavior demonstrated superior resistance to irradiation hardening compared to conventional austenitic stainless steel.

在奥氏体不锈钢中引入析出相是提高耐辐照性能的有效方法。本研究采用Fe、He离子对304高硼奥氏体不锈钢在室温下进行单独和同时辐照，研究了离子辐照对辐照缺陷和力学性能的影响机理。在不同辐照条件下，在整个离子穿透区，奥氏体和TiB2相均保持了原有的晶体结构，而（Fe,Cr）2B相则发生了晶向非晶转变。仅He离子辐照就产生了纳米级的氦气泡，并在奥氏体基体的损伤峰区诱发了高密度的精细位错环。在Fe+He离子同时辐照下，氦气泡的平均尺寸和氦气泡层宽度增大，同时Fe离子的掺入也促进了位错环的聚结和生长。离子辐照后（Fe,Cr）2B相的非晶化对软化起主导作用，而氦气泡的钉住作用则抑制了硬度的降低。辐照诱导的位错环在硬化中起主要作用，特别是仅He离子辐照诱导的小而致密的位错环产生了显著的硬化效应。在Fe+He离子同时照射下，硼化物弥散分布的304高硼不锈钢表现出轻微的软化，没有明显的硬化。与传统奥氏体不锈钢相比，该材料具有更好的抗辐照硬化性能。

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引用次数: 0

Combined effect of Te and impurities in molten LiF-NaF-KF salt on the corrosion of 316H SS and UNS N10003 alloy 熔盐中Te和杂质对316H SS和UNS N10003合金腐蚀的综合影响

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

Journal of Nuclear Materials

Pub Date : 2026-02-01 Epub Date: 2025-12-20 DOI: 10.1016/j.jnucmat.2025.156397

Miaomiao Hu , Xinmei Yang , Huajian Liu , Xingtai Zhou

The corrosion mechanisms of nickel-based alloy (UNS N10003) and stainless steel (316H SS) in molten LiF-NaF-KF (FLiNaK) salt at 700 °C were investigated. Results indicate that UNS N10003 exhibits better resistance to intergranular corrosion than 316H SS in the molten FLiNaK salt with and without Te. In the same batch of molten FLiNaK salt without Te, for UNS N10003 alloy, the reactions between its alloying elements and the impurities in the salt caused the uniform corrosion; while the reactions led to an intergranular corrosion for 316H SS. The presence of Te in molten FLiNaK salt induced the intergranular corrosion for UNS N10003 alloy. The coexistence of Te (1 wt% Te) and the impurities in the molten FLiNaK salt aggravated the corrosion of UNS N10003 alloy and 316H SS by ∼12 times. The corrosion depth of 316H SS (∼370 μm for 400 h) is larger than that of UNS N10003 (∼90 μm for 400 h) in the molten FLiNaK salt with Te (1 wt%). The severer intergranular corrosion is mainly attributed to the reactions involving Te, the impurities in salt, the alloying elements, and the precipitates at grain boundaries.

研究了镍基合金（UNS N10003）和不锈钢（316H SS）在700℃熔盐LiF-NaF-KF （FLiNaK）中的腐蚀机理。结果表明，UNS N10003在含Te和不含Te的熔融FLiNaK盐中均比316H SS具有更好的抗晶间腐蚀性能。在同一批次未添加Te的熔融FLiNaK盐中，对于UNS N10003合金，其合金元素与盐中的杂质发生反应，造成均匀腐蚀；熔盐中Te的存在引起了UNS N10003合金的晶间腐蚀。熔盐中Te （1 wt% Te）和杂质的共存使UNS N10003合金和316H SS的腐蚀加重了约12倍。316H SS在含Te （1 wt%）的熔融FLiNaK盐中的腐蚀深度（~ 370 μm, 400 h）大于UNS N10003 （~ 90 μm, 400 h）。严重的晶间腐蚀主要是由于Te、盐中的杂质、合金元素和晶界析出相的反应所致。

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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 : 2026-02-01 Epub 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。

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引用次数: 0

Performance of W and W-Re alloys under low-energy He+ ion irradiations at relevant nuclear fusion environments W和W- re合金在低能He+离子辐照下的核聚变性能

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

Journal of Nuclear Materials

Pub Date : 2026-01-01 Epub Date: 2025-11-03 DOI: 10.1016/j.jnucmat.2025.156278

Tyler E. Ray, Jitendra K. Tripathi, Jack C. Johnson, Ahmed Hassanein

Fusion-relevant tungsten (W) and tungsten-rhenium (W-Re) alloys were exposed to low-energy helium (He) ions at divertor base temperatures relevant to ITER/DEMO conditions. Pure W and W-Re alloys (1, 3, 5 wt. % Re) were irradiated with 100 eV He⁺ ions (fluence ∼

1.08 \times 10^{24}

ions m⁻²) from 900 to 1300 K. Extensive analysis utilized scanning electron microscopes (SEM) with focused ion beam (FIB) cross-sectioning, atomic force microscope (AFM), optical reflectivity, x-ray diffraction (XRD), and x-ray photoelectron spectroscopy (XPS) to investigate the surface evolution. Early-stage fuzz formed on all samples at 900 K, with bubble size increasing with Re content. At 1300 K, W-3Re and W-5Re exhibited surface roughening and pore formation, but no evidence of fuzz growth. Reflectivity increased significantly at 1300 K, with W-3Re reaching ∼47 %, higher than W-5Re, indicating less surface roughening with 3 wt. % Re. FIB cross-sectioning at 1100 K revealed a fuzz depth of ∼600 nm on both pure W and W-5Re samples, where under these experimental conditions of intermediate temperature, fuzz growth does not depend on Re concentration. X-ray diffractograms showed consistent crystallographic orientations for pure W and W-5Re. Microstrain analysis (Williamson-Hall) method revealed an increase of ∼0.070 ± 0.002 % in the annealed surface of W-5Re compared to pure W. Post-irradiation the microstrain increased significantly, with pure W exhibiting approximately threefold measured increase in microstrain compared to W-5Re, highlighting the stabilizing effect of substitutional Re. XPS results confirmed the Re content in the alloy and revealed an increased oxide percentage on more developed fuzzy surfaces. These results suggest potential mechanisms for W-Re alloys observed decrease in fuzz forming temperature window, providing crucial insights into future materials to potentially extend the divertor lifetime.

将与聚变相关的钨(W)和钨铼（W- re）合金暴露于低能氦（He）离子中，温度与ITER/DEMO条件相关。纯W和W-Re合金（1,3,5 wt. % Re）用100 eV He+离子（fluence ~ 1.08×1024 ions m−2）从900至1300 K辐照。利用聚焦离子束（FIB）的扫描电子显微镜（SEM）、原子力显微镜（AFM）、光学反射率、x射线衍射（XRD）和x射线光电子能谱（XPS）对表面演化进行了广泛的分析。在900 K时，所有样品都形成了早期的绒毛，气泡大小随Re含量的增加而增加。在1300 K时，W-3Re和W-5Re表现出表面粗化和孔隙形成，但没有绒毛生长的迹象。在1300 K时，反射率显著增加，W- 3re达到~ 47%，高于W- 5re，这表明在3 wt. % Re下，表面粗化程度较低。1100 K时FIB横截面显示，纯W和W- 5re样品的绒毛深度均为~ 600 nm，其中在这些中间温度的实验条件下，绒毛的生长与Re浓度无关。x射线衍射图显示纯W和W- 5re的晶体取向一致。微应变分析（Williamson-Hall）方法显示，W- 5re退火表面的微应变比纯W增加了~ 0.070±0.002%。辐照后，微应变显著增加，纯W的微应变比W- 5re增加了约三倍，突出了取代Re的稳定作用。XPS结果证实了合金中的Re含量，并显示在更发育的模糊表面上增加了氧化物百分比。这些结果为W-Re合金观察到的模糊形成温度窗的减少提供了潜在的机制，为未来材料提供了重要的见解，以潜在地延长导流器的寿命。

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引用次数: 0

Interstitial-mediated precipitation unveiled: Mechanism of irradiation-induced rhenium aggregation in molybdenum 间质沉淀揭示：辐照诱导铼在钼中的聚集机制

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

Journal of Nuclear Materials

Pub Date : 2026-01-01 Epub Date: 2025-11-05 DOI: 10.1016/j.jnucmat.2025.156283

Xiang-Qian Wu , Wei Zhao , Yu-Hao Li , Zhi-Peng Sun , Yong Xin , Xi Qiu , Huiqiu Deng , Hong-Bo Zhou , Guang-Hong Lu

We investigated the mechanisms of rhenium (Re) aggregation and precipitation in molybdenum (Mo) under irradiation by employing a combination of first-principles calculations and object kinetic Monte Carlo (OKMC) simulations. Our results demonstrate that irradiation-induced defects significantly affect the distribution of Re, resulting in distinct patterns compared to non-irradiated conditions. In pristine Mo, Re atoms favor the even distribution due to the repulsive interactions. In contrast, strong attractive interactions are observed between Re and defects, particularly self-interstitial atoms (SIAs). Furthermore, the rapid three-dimensional diffusion of Re-SIA complexes in Mo facilitates their efficient trapping by substitutional Re atoms, leading to the formation of stable Re-Re dumbbells. These dumbbells subsequently serve as effective traps for subsequent Re atoms, thereby driving the nucleation and growth of Re-rich clusters. Based on these findings, we propose an interstitial-mediated migration and aggregation mechanism for the irradiation-induced Re precipitation in Mo. Building on these insights, we utilized OKMC simulations to examine the co-evolution of Re and irradiation-induced defects in Mo. The simulations reveal that Re atoms aggregate to form Re-rich clusters, and eventually precipitates, with increasing irradiation dose. Moreover, the number of Re atoms in interstitial-type clusters surpasses that in vacancy-type clusters, corroborating the interstitial-mediated mechanism. These results provide a valuable reference for the composition optimization and property evaluation of Mo-Re alloys under irradiation.

采用第一性原理计算和物体动力学蒙特卡罗（OKMC）模拟相结合的方法研究了辐照下钼（Mo）中铼（Re）的聚集和沉淀机理。我们的研究结果表明，辐照诱导缺陷显著影响Re的分布，导致与未辐照条件相比不同的模式。在原始Mo中，由于排斥相互作用，Re原子倾向于均匀分布。相反，Re和缺陷，特别是自间隙原子（SIAs）之间观察到强烈的吸引相互作用。此外，Re- sia配合物在Mo中的快速三维扩散有助于它们被取代的Re原子有效捕获，从而形成稳定的Re-Re哑铃。这些哑铃随后成为后续Re原子的有效陷阱，从而驱动富Re簇的成核和生长。基于这些发现，我们提出了辐照诱导Mo中Re沉淀的间隙介导迁移和聚集机制。基于这些发现，我们利用OKMC模拟来研究Mo中Re和辐照诱导缺陷的共同演化。模拟显示，随着辐照剂量的增加，Re原子聚集形成富Re簇，最终沉淀。此外，间隙型团簇中的Re原子数量超过了空位型团簇，证实了间隙介导的机制。这些结果为辐照下Mo-Re合金的成分优化和性能评价提供了有价值的参考。

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引用次数: 0

Atomistic insights into dislocation–void interactions in tungsten–vanadium alloys under fusion-Relevant conditions 熔合相关条件下钨钒合金中位错-空洞相互作用的原子见解

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

Journal of Nuclear Materials

Pub Date : 2026-01-01 Epub Date: 2025-11-29 DOI: 10.1016/j.jnucmat.2025.156319

Ayobami D. Daramola

Tungsten–vanadium (W–V) alloys are promising candidates for fusion plasma-facing components, yet irradiation-induced nanoscale voids can strongly hinder dislocation glide and degrade mechanical performance. Here we use large-scale molecular-dynamics simulations to examine edge-dislocation motion and void bypass in single-crystal W and random W–V alloys containing 5 and 10 at.% V over a fusion-relevant temperature range (300–900 K) and void diameters of 2–5 nm. Vanadium reduces the unstable stacking-fault energy anisotropically while preserving high-temperature stiffness, and introduces substantial solute drag that elevates the CRSS for obstacle-free glide, particularly at intermediate temperatures. In the presence of voids, bypass occurs through a reproducible sequence involving screw-segment nucleation, prismatic loop emission, Orowan bow-out, and jog-mediated release. The stresses associated with these stages increase with V content and diminish with temperature. The peak bypass stress, τ_max(d_eff, T), is described accurately by a modified Bacon–Kocks–Scattergood line-tension formulation in which geometric scaling is retained and solute and thermal effects enter as nearly additive offsets. The resulting obstacle strengths, α, for 2 nm voids compare favourably with recent in-situ TEM measurements at 293 and 723 K. Collectively, these results provide stage-resolved mechanistic insight, deliver continuum-ready hardening parameters, and identify  ∼ 10 at.% V as a composition that offers a favourable balance between enhanced void resistance and preserved post-bypass mobility in tungsten-based fusion materials.

钨钒（W-V）合金是有前途的候选融合等离子体组件，但辐照诱导的纳米级空洞会严重阻碍位错滑动和降低机械性能。本文采用大规模分子动力学模拟研究了含5和10 at的单晶W和随机W - v合金中的边位错运动和空穴旁路。在熔合相关温度范围（300-900 K）和空隙直径为2-5 nm的范围内。钒在保持高温刚度的同时，降低了不稳定的堆叠断层能量的各向异性，并引入了大量的溶质阻力，提高了CRSS的无障碍滑行，特别是在中等温度下。在存在空隙的情况下，旁路通过一个可重复的序列发生，包括螺旋段成核、棱柱形环发射、Orowan弓出和慢跑介导的释放。这些阶段的应力随V含量的增加而增大，随温度的升高而减小。峰值旁路应力τmax（deff， T）可以用修正的Bacon-Kocks-Scattergood线张力公式精确地描述，其中保留了几何标度，溶质和热效应作为几乎可加的偏移量进入。在293和723 K下，得到的2 nm孔洞的障碍强度α与最近的原位透射电镜测量结果比较有利。总的来说，这些结果提供了阶段分解机制的洞察力，提供了连续硬化参数，并确定了 ~ 10 at。% V作为一种组合物，在钨基熔合材料中提供了增强的空隙电阻和保留的旁路迁移率之间的良好平衡。

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引用次数: 0

Investigation of the incorporation capacity of calcine of simulated high-level liquid waste in borosilicate glass matrix 模拟高放废液煅烧在硼硅酸盐玻璃基体中的掺入能力研究

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

Journal of Nuclear Materials

Pub Date : 2026-01-01 Epub Date: 2025-10-05 DOI: 10.1016/j.jnucmat.2025.156208

Wang Changfu , Liu Xu , Li Lin , Li Wengzheng , Shao Bobo , Liu Yan , Wang Yun , Huang Xi , Tan Shengheng , Liu Zhirong , Zhang Shengdong

As a pivotal technology for high-level liquid waste (HLW) treatment, vitrification plays a crucial role in ensuring the sustainable development of nuclear energy. The two-step vitrification process, which involves initial calcination of HLW into solid calcine followed by melting with base glass, has demonstrated significant improvements in immobilization efficiency. This study investigated the capacity of simulated HLW calcine to be incorporated within a borosilicate glass matrix.Through comprehensive characterization techniques including field emission scanning electron microscopy coupled with energy dispersive spectroscopy (SEM-EDS), X-ray diffraction (XRD), Fourier-transform infrared spectroscopy (FTIR), and Raman spectroscopy, the effects of calcine content (18–32 wt%) on the microstructure, chemical composition, phase distribution, and network structure of the vitrified matrixs were examined. Homogeneous glass matrices are achieved at 18–26 wt% calcine content, while phase separation and crystalline precipitation occur at higher concentrations. However, when the content reaches 28 wt%, the initial precipitation of granular CaMoO₄ crystals occurs, maintaining relatively uniform distribution. Further increasing the content to 30 wt% leads to the formation of acicular CaMoO₄ phases accompanied by minor Zr_xCe_1-xO₂ particle precipitation. Notably, at 32 wt% calcined product content, a distinct phase separation layer emerges on the surface of the vitrified matrix, with a substantial increase in Zr_xCe_1-xO₂ precipitates. These research findings have clarified the incorporation capacity limits of simulated HLW calcine in borosilicate glass matrix, providing valuable theoretical guidance for enhancing waste loading in vitrified forms during the vitrification process.

玻璃化处理作为高放废液处理的关键技术，对保证核能的可持续发展起着至关重要的作用。两步玻璃化工艺，包括将高废渣初始煅烧成固体煅烧，然后与基玻璃熔化，已经证明了固定效率的显着提高。本研究考察了模拟高锰酸盐煅烧物在硼硅酸盐玻璃基体中的掺入能力。通过场发射扫描电子显微镜（SEM-EDS）、x射线衍射（XRD）、傅里叶变换红外光谱（FTIR）和拉曼光谱等综合表征技术，研究了煅烧量（18-32 wt%）对玻璃化基体微观结构、化学成分、相分布和网络结构的影响。当煅烧物含量为18-26 wt%时，可以实现均匀的玻璃基质，而当煅烧物浓度较高时，会发生相分离和结晶析出。而当添加量达到28 wt%时，CaMoO4颗粒状晶体开始析出，并保持相对均匀的分布。进一步增加到30 wt%，形成针状CaMoO4相，并伴有少量ZrxCe1-xO2颗粒析出。值得注意的是，当煅烧产物含量为32 wt%时，玻璃化基体表面出现明显的相分离层，ZrxCe1-xO2析出物大量增加。这些研究结果明确了模拟高沸石煅烧料在硼硅玻璃基体中的掺入容量限制，为在玻璃化过程中提高玻璃化形式的废物装载量提供了有价值的理论指导。

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引用次数: 0

Revealing the superior SCC resistance of a dual-phase stainless steel in simulated PWR primary water via high-resolution characterization 通过高分辨率表征揭示了双相不锈钢在模拟压水堆一次水中优越的SCC抗性

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

Journal of Nuclear Materials

Pub Date : 2026-01-01 Epub Date: 2025-11-20 DOI: 10.1016/j.jnucmat.2025.156317

Jun Xiao , Yuhao Zhou , Ting Xiao , Cheng Peng , Hao Wang , Jiamei Wang , Kai Chen

The stress corrosion cracking (SCC) resistance of a Z3CN20–09 M duplex stainless steel (DSS) was systematically studied and compared to conventional single-phase austenitic stainless steel (316NG) under simulated pressurized water reactor (PWR) primary water conditions. Results show that the DSS exhibits much lower SCC growth rates than the 316NG in either deformed or high corrosion potential conditions. High-resolution microstructural characterization revealed the oxidation and crack-tip behaviors during SCC tests in simulated primary water, while in-situ SEM tensile experiments elucidated the deformation compatibility between austenite and ferrite phases, confirming that the duplex structure fundamentally enhances SCC resistance. The ferrite phase, which contains higher Cr, appears to promote the formation of protective chromium-rich oxide layers, thereby reducing oxidation-driven crack propagation. Additionally, ferrite shows lower strain accumulation and appears to limit crack propagation at austenite-ferrite interfaces.

系统研究了Z3CN20-09 M双相不锈钢（DSS）在模拟压水堆（PWR）一次水条件下的抗应力腐蚀开裂（SCC）性能，并与常规单相奥氏体不锈钢（316NG）进行了比较。结果表明，在变形或高腐蚀电位条件下，DSS的SCC生长速率都比316NG低得多。高分辨率显微组织表征揭示了模拟原生水中SCC试验中的氧化和裂纹尖端行为，而原位SEM拉伸实验揭示了奥氏体和铁素体相之间的变形相容性，证实了双相结构从根本上增强了抗SCC能力。铁素体相含有较高的Cr，促进了保护性富铬氧化层的形成，从而减少了氧化裂纹扩展。此外，铁素体表现出较低的应变积累，并在奥氏体-铁素体界面上限制裂纹扩展。

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引用次数: 0