Pub Date : 2025-12-13DOI: 10.1016/j.jnucmat.2025.156377
Hongxia Xu , Bin Leng , Qi Liu , Wei Xu , Jiandang Liu , Bangjiao Ye , Hefei Huang , Xingtai Zhou , Run Ye
Positron annihilation spectroscopy reveals the effect of tellurium on the corrosion behavior of UNS N10003 in FLiNaK salts at 650°C. Te exposure increased mass loss 16-fold (0.48 vs 0.03 mg/cm²), with inductively coupled plasma optical emission spectrometry (ICP-OES) showing elevated Fe and Cr concentrations but depleted Te in the FLiNaK salts. X-ray diffraction (XRD) and scanning electron microscopy (SEM) identified Ni₃Te₂ precipitates at grain boundaries (GBs) and M6C carbide-matrix interfaces, which induce the τ₃ component in positron annihilation spectra due to void trapping. Coincidence Doppler broadening (CDB) confirmed defect proliferation via elevation of the S-parameter. Ni₃Te₂ precipitates failed to suppress Cr dissolution and accelerated GBs weakening, thereby promoting intergranular cracking or embrittlement.
正电子湮没光谱揭示了碲对UNS N10003在650℃下在FLiNaK盐中的腐蚀行为的影响。暴露使质量损失增加了16倍（0.48 vs 0.03 mg/cm²），电感耦合等离子体光学发射光谱（ICP-OES）显示，FLiNaK盐中的Fe和Cr浓度升高，但Te耗尽。x射线衍射（XRD）和扫描电镜（SEM）在晶界（GBs）和M6C碳化物-基体界面处发现Ni₃Te₂沉淀，由于空穴捕获导致正电子湮没谱中出现τ₃分量。重合多普勒增宽（CDB）通过s参数升高证实了缺陷增生。Ni₃Te₂的析出未能抑制Cr的溶解，反而加速了GBs的弱化，从而促进了晶间开裂或脆化。
{"title":"Poisitron annihilation study on corrosion-incudced defects in UNS N10003 alloy exposed to tellurium-containing FLiNaK salt","authors":"Hongxia Xu , Bin Leng , Qi Liu , Wei Xu , Jiandang Liu , Bangjiao Ye , Hefei Huang , Xingtai Zhou , Run Ye","doi":"10.1016/j.jnucmat.2025.156377","DOIUrl":"10.1016/j.jnucmat.2025.156377","url":null,"abstract":"<div><div>Positron annihilation spectroscopy reveals the effect of tellurium on the corrosion behavior of UNS N10003 in FLiNaK salts at 650°C. Te exposure increased mass loss 16-fold (0.48 vs 0.03 mg/cm²), with inductively coupled plasma optical emission spectrometry (ICP-OES) showing elevated Fe and Cr concentrations but depleted Te in the FLiNaK salts. X-ray diffraction (XRD) and scanning electron microscopy (SEM) identified Ni₃Te₂ precipitates at grain boundaries (GBs) and M<sub>6</sub>C carbide-matrix interfaces, which induce the τ₃ component in positron annihilation spectra due to void trapping. Coincidence Doppler broadening (CDB) confirmed defect proliferation via elevation of the S-parameter. Ni₃Te₂ precipitates failed to suppress Cr dissolution and accelerated GBs weakening, thereby promoting intergranular cracking or embrittlement.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156377"},"PeriodicalIF":3.2,"publicationDate":"2025-12-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789294","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}
Pub Date : 2025-12-13DOI: 10.1016/j.jnucmat.2025.156380
Jiale Huang , Jintao Zhang , Chenfei Cui , Xueli Mao , Xiaodan Fei , Yang Wu , Guo Pu , Haoxuan Zhong , Sijie Liu , Fangfang Ge , Bingsheng Li
This study focused on the early-stage corrosion behavior of the Al2O3 coatings. Al2O3 coatings were deposited on SIMP steel substrates via magnetron sputtering. Subsequent corrosion experiments were conducted in oxygen-saturated lead-bismuth eutectic (LBE) at 500 °C for 300 h and 500 h, and 600 °C for 300 h, with complementary testing under an oxygen-controlled condition (10−6 wt.%) at 500 °C for 500 h. The coating exhibited an initial thickness increase after 300 h at 500 °C in oxygen-saturated LBE, followed by a decrease after 500 h. In contrast, under the oxygen-controlled condition, the coating thickness remained nearly unchanged throughout the 500 h exposure. This divergence is attributed to the initial formation and decomposition of PbAl2O4 in oxygen-saturated environment, whereas formation of PbAl2O4 was suppressed under oxygen-controlled condition. The coating’s maintained corrosion resistance despite phase changes confirms the efficacy of oxygen control and its promise for structural material applications.
{"title":"Early-stage corrosion behavior of amorphous Al2O3 coating on SIMP steel in static lead-bismuth eutectic","authors":"Jiale Huang , Jintao Zhang , Chenfei Cui , Xueli Mao , Xiaodan Fei , Yang Wu , Guo Pu , Haoxuan Zhong , Sijie Liu , Fangfang Ge , Bingsheng Li","doi":"10.1016/j.jnucmat.2025.156380","DOIUrl":"10.1016/j.jnucmat.2025.156380","url":null,"abstract":"<div><div>This study focused on the early-stage corrosion behavior of the Al<sub>2</sub>O<sub>3</sub> coatings. Al<sub>2</sub>O<sub>3</sub> coatings were deposited on SIMP steel substrates via magnetron sputtering. Subsequent corrosion experiments were conducted in oxygen-saturated lead-bismuth eutectic (LBE) at 500 °C for 300 h and 500 h, and 600 °C for 300 h, with complementary testing under an oxygen-controlled condition (10<sup>−6</sup> wt.%) at 500 °C for 500 h. The coating exhibited an initial thickness increase after 300 h at 500 °C in oxygen-saturated LBE, followed by a decrease after 500 h. In contrast, under the oxygen-controlled condition, the coating thickness remained nearly unchanged throughout the 500 h exposure. This divergence is attributed to the initial formation and decomposition of PbAl<sub>2</sub>O<sub>4</sub> in oxygen-saturated environment, whereas formation of PbAl<sub>2</sub>O<sub>4</sub> was suppressed under oxygen-controlled condition. The coating’s maintained corrosion resistance despite phase changes confirms the efficacy of oxygen control and its promise for structural material applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156380"},"PeriodicalIF":3.2,"publicationDate":"2025-12-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145754016","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}
Pub Date : 2025-12-13DOI: 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.
{"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}
Pub Date : 2025-12-11DOI: 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.
{"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}
Pub Date : 2025-12-11DOI: 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.
{"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}
Pub Date : 2025-12-11DOI: 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.
{"title":"First experimental determination of the low-temperature isothermal section of the Zr-Nb-Cr ternary system at 1073 K","authors":"Wen-bo Wang , Kang Wang , Wenfang Li , Jun Du","doi":"10.1016/j.jnucmat.2025.156367","DOIUrl":"10.1016/j.jnucmat.2025.156367","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156367"},"PeriodicalIF":3.2,"publicationDate":"2025-12-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838045","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}
Pub Date : 2025-12-11DOI: 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 %).
{"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}
Pub Date : 2025-12-11DOI: 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.
{"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}
Pub Date : 2025-12-10DOI: 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.
{"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}
Pub Date : 2025-12-10DOI: 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.
{"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}
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