Pub Date : 2026-03-01Epub Date: 2025-12-25DOI: 10.1016/j.jnucmat.2025.156413
YanBang Tang
Predicting the onset dose of void swelling is a critical challenge in developing radiation-resistant materials, a task often hindered by sparse and heterogeneous experimental data. To address this, we present a comprehensive framework combining a Gaussian noise-based data augmentation strategy with the AutoGluon automated machine learning (AutoML) platform. This study introduces a newly expanded dataset, updated with 80 recent publications (2020-2025) to form a comprehensive library of 374 irradiated metal samples. Our proposed framework's efficacy is rigorously evaluated by applying the augmentation strategy to AutoGluon, as well as to two state-of-the-art (SOTA) tabular models, TabM and TabPFN. The augmented AutoGluon model demonstrated superior performance, achieving a Root Mean Squared Error (RMSE) of 23.19 dpa and a coefficient of determination (R²) of 0.872 on an unseen test set. This represents a 6.3 % reduction in error compared to its baseline and outperforms the augmented SOTA models. The results consistently show that data augmentation improves performance across all model architectures. SHapley Additive exPlanations (SHAP) analysis of the superior model confirmed its physical interpretability, identifying key features and their complex interactions. This synergistic methodology demonstrates a powerful, validated pathway to overcome data scarcity in materials informatics, enhancing predictive power and accelerating the data-driven design of advanced alloys.
{"title":"Automated machine learning with data augmentation for predicting void swelling onset dose in irradiated metals","authors":"YanBang Tang","doi":"10.1016/j.jnucmat.2025.156413","DOIUrl":"10.1016/j.jnucmat.2025.156413","url":null,"abstract":"<div><div>Predicting the onset dose of void swelling is a critical challenge in developing radiation-resistant materials, a task often hindered by sparse and heterogeneous experimental data. To address this, we present a comprehensive framework combining a Gaussian noise-based data augmentation strategy with the AutoGluon automated machine learning (AutoML) platform. This study introduces a newly expanded dataset, updated with 80 recent publications (2020-2025) to form a comprehensive library of 374 irradiated metal samples. Our proposed framework's efficacy is rigorously evaluated by applying the augmentation strategy to AutoGluon, as well as to two state-of-the-art (SOTA) tabular models, TabM and TabPFN. The augmented AutoGluon model demonstrated superior performance, achieving a Root Mean Squared Error (RMSE) of 23.19 dpa and a coefficient of determination (R²) of 0.872 on an unseen test set. This represents a 6.3 % reduction in error compared to its baseline and outperforms the augmented SOTA models. The results consistently show that data augmentation improves performance across all model architectures. SHapley Additive exPlanations (SHAP) analysis of the superior model confirmed its physical interpretability, identifying key features and their complex interactions. This synergistic methodology demonstrates a powerful, validated pathway to overcome data scarcity in materials informatics, enhancing predictive power and accelerating the data-driven design of advanced alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156413"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145845597","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 : 2026-03-01Epub Date: 2026-01-09DOI: 10.1016/j.jnucmat.2026.156443
Shuang Hu , Zhen Wu , Yao Yu , Mei Zhou , Qigui Yang , Peng Zhang , Yu Chen , Mingpan Wan , Te Zhu , Xingzhong Cao
Although titanium alloys have gained significant attention for their potential applications in advanced reactors, experimental studies on irradiation damage under varied irradiation conditions remain insufficient, limiting the understanding of defect evolution and hardening behavior. This study selected the near α titanium alloy Ti-5Al-3V-3Zr-Cr (Ti-5331), which has ideal mechanical properties, and compares its irradiation-induced defect formation, softening, and hardening effects under different fluences and temperatures. Results from slow positron-beam Doppler broadening spectroscopy (DBS) confirm that hydrogen ion irradiation generates a significant number of vacancy-type defects and HmVn complexes in room temperature (RT) and high temperatures (473 K and 573 K). At a high fluence (RT-1 × 1017 H⁺/cm²), the excess HmVn complexes will inhibit the increase in the S parameter. In contrast, at 473 K and 573 K, thermal activation reduces the concentration of vacancy-type defects, and led to a significant decrease in the overall S parameter. In addition to the aforementioned defects, a large number of hydrogen atoms occupying vacancies gradually form small hydrogen bubbles, which increase in size with increasing fluence (5 × 1016 H⁺/cm² to 1 × 1017 H⁺/cm²) and temperature (RT to 573 K). Notably, the hydrogen bubbles in the α phase are larger than those in the β phase (e.g. RT-1 × 1017 H⁺/cm² sample). Unlike the typical irradiation hardening phenomenon, the nanoindentation results exhibit significant irradiation softening. The softening effect becomes more pronounced with increasing room-temperature irradiation fluence, resulting in a hardness reduction of up to 19% compared to the unirradiated samples. Irradiation at elevated temperatures also resulted in significant softening. The softening effect may be attributed to hydrogen-induced local plastic deformation, where hydrogen enhances the interaction of dislocations on different slip planes, leading to the increased complexity of dislocation structures and increased local plasticity. These findings elucidate hydrogen-defect interactions and temperature-fluence synergies, critical for designing irradiation-resistant titanium alloys in nuclear applications.
{"title":"Hydrogen ion irradiation-induced defect evolution and softening in near-α Ti-5331 alloy: Effects of fluence and temperature","authors":"Shuang Hu , Zhen Wu , Yao Yu , Mei Zhou , Qigui Yang , Peng Zhang , Yu Chen , Mingpan Wan , Te Zhu , Xingzhong Cao","doi":"10.1016/j.jnucmat.2026.156443","DOIUrl":"10.1016/j.jnucmat.2026.156443","url":null,"abstract":"<div><div>Although titanium alloys have gained significant attention for their potential applications in advanced reactors, experimental studies on irradiation damage under varied irradiation conditions remain insufficient, limiting the understanding of defect evolution and hardening behavior. This study selected the near α titanium alloy Ti-5Al-3V-3Zr-Cr (Ti-5331), which has ideal mechanical properties, and compares its irradiation-induced defect formation, softening, and hardening effects under different fluences and temperatures. Results from slow positron-beam Doppler broadening spectroscopy (DBS) confirm that hydrogen ion irradiation generates a significant number of vacancy-type defects and H<sub>m</sub>V<sub>n</sub> complexes in room temperature (RT) and high temperatures (473 K and 573 K). At a high fluence (RT-1 × 10<sup>17</sup> H⁺/cm²), the excess H<sub>m</sub>V<sub>n</sub> complexes will inhibit the increase in the S parameter. In contrast, at 473 K and 573 K, thermal activation reduces the concentration of vacancy-type defects, and led to a significant decrease in the overall S parameter. In addition to the aforementioned defects, a large number of hydrogen atoms occupying vacancies gradually form small hydrogen bubbles, which increase in size with increasing fluence (5 × 10<sup>16</sup> H⁺/cm² to 1 × 10<sup>17</sup> H⁺/cm²) and temperature (RT to 573 K). Notably, the hydrogen bubbles in the α phase are larger than those in the β phase (e.g. RT-1 × 10<sup>17</sup> H⁺/cm² sample). Unlike the typical irradiation hardening phenomenon, the nanoindentation results exhibit significant irradiation softening. The softening effect becomes more pronounced with increasing room-temperature irradiation fluence, resulting in a hardness reduction of up to 19% compared to the unirradiated samples. Irradiation at elevated temperatures also resulted in significant softening. The softening effect may be attributed to hydrogen-induced local plastic deformation, where hydrogen enhances the interaction of dislocations on different slip planes, leading to the increased complexity of dislocation structures and increased local plasticity. These findings elucidate hydrogen-defect interactions and temperature-fluence synergies, critical for designing irradiation-resistant titanium alloys in nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156443"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024745","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}
Metal tritides have long been applied in tritium storage due to their high capacity and stability. The decay of tritium produces helium-3 (³He), which is mainly retained in metal tritides in the form of bubbles. Although the evolution of helium-3 bubbles in metal tritides has been of wide concern for a long time, the trend of their morphological transformation is still under debate. In this work, the shape evolution of helium bubbles in typical metal tritides (erbium, titanium, and zirconium) was tracked by transmission electron microscopy. The results show that in the tritides of erbium and titanium, helium-3 bubbles undergo a sphere-to-platelet transformation at the early stage (³He/M = 0.02∼0.06), while in zirconium tritide the helium-3 bubbles remain spherical up to ³He/M > 0.29. Compared with theoretical models, it is found that large and plate-like bubbles can maintain stability by widening rather than through a spherical transformation. Our results further suggest that the dominant energy contribution of helium-3 bubbles shifts from surface energy to strain energy with aging. Overall, the present work provides strong experimental support for investigating helium bubble behavior within metal tritide lattices, offering guidance for the rational design of tritium storage and fusion materials.
{"title":"Spontaneous shape transformation of helium bubble in metal tritide lattice: sphere to platelet","authors":"Muhong Li, Lin Qi, Chengqin Zou, Shuanglin Hu, Weidu Wang, Xiaochun Han, Xiaosong Zhou, Shuming Peng, Huahai Shen","doi":"10.1016/j.jnucmat.2026.156470","DOIUrl":"10.1016/j.jnucmat.2026.156470","url":null,"abstract":"<div><div>Metal tritides have long been applied in tritium storage due to their high capacity and stability. The decay of tritium produces helium-3 (³He), which is mainly retained in metal tritides in the form of bubbles. Although the evolution of helium-3 bubbles in metal tritides has been of wide concern for a long time, the trend of their morphological transformation is still under debate. In this work, the shape evolution of helium bubbles in typical metal tritides (erbium, titanium, and zirconium) was tracked by transmission electron microscopy. The results show that in the tritides of erbium and titanium, helium-3 bubbles undergo a sphere-to-platelet transformation at the early stage (³He/M = 0.02∼0.06), while in zirconium tritide the helium-3 bubbles remain spherical up to ³He/M > 0.29. Compared with theoretical models, it is found that large and plate-like bubbles can maintain stability by widening rather than through a spherical transformation. Our results further suggest that the dominant energy contribution of helium-3 bubbles shifts from surface energy to strain energy with aging. Overall, the present work provides strong experimental support for investigating helium bubble behavior within metal tritide lattices, offering guidance for the rational design of tritium storage and fusion materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156470"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024751","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 : 2026-03-01Epub Date: 2026-01-17DOI: 10.1016/j.jnucmat.2026.156467
Elina Charatsidou , Anita Pazzaglia , Kaitlyn Bullock , Maria Giamouridou , Eleanor Lawrence Bright , Mikael Jolkkonen , Christoph Hennig , Pär Olsson
Uranium mononitride (UN) is a promising candidate fuel for next-generation fast reactors due to its high fissile density, superior thermal conductivity, and high melting point compared to conventional oxide fuels. However, scarce experimental data on UN and its thermophysical behaviour under fission product incorporation limits its performance assessment. Zirconium nitride (ZrN) is an efficient thermal conductor and a candidate material for inert matrix fuels. Given its high thermal conductivity, ZrN addition at sufficient concentrations should, in principle, induce percolation conduction and increase thermal conductivity in UN. To decouple chemistry from irradiation-induced porosity, known to dominate thermal degradation at high burnup, this study isolates the intrinsic chemical contribution of Zr incorporation under dense, low-porosity conditions. (U,Zr)N pellets with 6.5 and 20 at. % Zr were fabricated by spark plasma sintering (SPS), using powders produced from arc-melted alloy via the hydride-nitride-denitride route. Synchrotron powder X-ray diffraction confirmed the formation of solid solutions and enhanced Zr solubility after sintering, resulting in improved microstructural homogeneity. Thermal diffusivity was measured between 300 and 1500 K using light flash analysis, and thermal conductivity was derived using heat capacity and density correlations with porosity correction. Despite the intrinsically higher thermal conductivity of ZrN, the incorporation of 6.5 at. % Zr reduced the thermal conductivity relative to UN, consistent with impurity scattering. The 20 at. % Zr composition further decreased conductivity, indicating the microstructure does not meet the conditions required for percolation conduction. Differences in the temperature dependence of thermal diffusivity between UN and Zr-bearing samples highlight a compositional influence on heat transport. The results provide benchmark data for (U,Zr)N and insights into chemical and thermophysical interactions in nitride ceramics.
{"title":"Impact of zirconium incorporation on the thermophysical properties of uranium mononitride","authors":"Elina Charatsidou , Anita Pazzaglia , Kaitlyn Bullock , Maria Giamouridou , Eleanor Lawrence Bright , Mikael Jolkkonen , Christoph Hennig , Pär Olsson","doi":"10.1016/j.jnucmat.2026.156467","DOIUrl":"10.1016/j.jnucmat.2026.156467","url":null,"abstract":"<div><div>Uranium mononitride (UN) is a promising candidate fuel for next-generation fast reactors due to its high fissile density, superior thermal conductivity, and high melting point compared to conventional oxide fuels. However, scarce experimental data on UN and its thermophysical behaviour under fission product incorporation limits its performance assessment. Zirconium nitride (ZrN) is an efficient thermal conductor and a candidate material for inert matrix fuels. Given its high thermal conductivity, ZrN addition at sufficient concentrations should, in principle, induce percolation conduction and increase thermal conductivity in UN. To decouple chemistry from irradiation-induced porosity, known to dominate thermal degradation at high burnup, this study isolates the intrinsic chemical contribution of Zr incorporation under dense, low-porosity conditions. (U,Zr)N pellets with 6.5 and 20 at. % Zr were fabricated by spark plasma sintering (SPS), using powders produced from arc-melted alloy via the hydride-nitride-denitride route. Synchrotron powder X-ray diffraction confirmed the formation of solid solutions and enhanced Zr solubility after sintering, resulting in improved microstructural homogeneity. Thermal diffusivity was measured between 300 and 1500 K using light flash analysis, and thermal conductivity was derived using heat capacity and density correlations with porosity correction. Despite the intrinsically higher thermal conductivity of ZrN, the incorporation of 6.5 at. % Zr reduced the thermal conductivity relative to UN, consistent with impurity scattering. The 20 at. % Zr composition further decreased conductivity, indicating the microstructure does not meet the conditions required for percolation conduction. Differences in the temperature dependence of thermal diffusivity between UN and Zr-bearing samples highlight a compositional influence on heat transport. The results provide benchmark data for (U,Zr)N and insights into chemical and thermophysical interactions in nitride ceramics.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156467"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024752","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 : 2026-03-01Epub Date: 2025-12-26DOI: 10.1016/j.jnucmat.2025.156416
Calum S. Cunningham, Georgios Papanikos
Predicting irradiation-induced embrittlement of a nuclear reactor pressure vessel (RPV) is paramount to safe operation. Machine learning (ML) is an emerging statistical analysis tool with the potential to improve on conventional predictive models. Developing a supervised ML model requires many user-dependent decisions during data processing, not least the selection of a database, many of which differ amongst past studies. This work investigates a range of different procedures for developing an ML model to predict the transition temperature shift (ΔT41J), with the aim of identifying the best approaches to support the generation of future best practice guidelines. The impacts of different pre-processing and data selection approaches are evaluated, including comparison with conventional models and particular focus on the ability of ML models to extrapolate beyond the training data domain. A multi-layer perceptron ensemble model is found to be optimal and capable of producing more accurate predictions than a leading embrittlement trend curve. It is shown that differences in pre processing choices are not highly influential on final models. However, training data selection is critically important due to the unreliability of ML models when extrapolating. Recommendations are provided to maximise the benefit of ML in predicting RPV structural integrity and to support the adoption of safe, reliable best practice guidance in codes and standards.
{"title":"Using machine learning to predict reactor pressure vessel embrittlement: Human factors and best practice","authors":"Calum S. Cunningham, Georgios Papanikos","doi":"10.1016/j.jnucmat.2025.156416","DOIUrl":"10.1016/j.jnucmat.2025.156416","url":null,"abstract":"<div><div>Predicting irradiation-induced embrittlement of a nuclear reactor pressure vessel (RPV) is paramount to safe operation. Machine learning (ML) is an emerging statistical analysis tool with the potential to improve on conventional predictive models. Developing a supervised ML model requires many user-dependent decisions during data processing, not least the selection of a database, many of which differ amongst past studies. This work investigates a range of different procedures for developing an ML model to predict the transition temperature shift (ΔT<sub>41J</sub>), with the aim of identifying the best approaches to support the generation of future best practice guidelines. The impacts of different pre-processing and data selection approaches are evaluated, including comparison with conventional models and particular focus on the ability of ML models to extrapolate beyond the training data domain. A multi-layer perceptron ensemble model is found to be optimal and capable of producing more accurate predictions than a leading embrittlement trend curve. It is shown that differences in pre processing choices are not highly influential on final models. However, training data selection is critically important due to the unreliability of ML models when extrapolating. Recommendations are provided to maximise the benefit of ML in predicting RPV structural integrity and to support the adoption of safe, reliable best practice guidance in codes and standards.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156416"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145957738","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 : 2026-03-01Epub Date: 2026-01-05DOI: 10.1016/j.jnucmat.2026.156440
Fihan Alharbi, Hywel Ragnauth, Timothy Abram, Joel Turner
Uranium dioxide (UO₂) is the standard nuclear fuel for light water reactors (LWRs) due to its operational experience, irradiation stability, and ease of manufacture. However, its low thermal conductivity results in a high temperature gradient across the pellet in-service, leading to thermal stresses, deformation, and cracking. This study explores the addition of a high thermal conductivity molybdenum phase as a nano-powder, which is trialled alongside pre-sintered UO2 granules to improve the thermal conductivity of the resulting pellet. UO2- 10 wt. % Mo composite pellets were fabricated by dispersing UO₂ granules and nano-Mo powder in ethanol during mixing, followed by ethanol evaporation and subsequent Spark Plasma Sintering (SPS) at 1473 K with a 5-minute hold. Pellet microstructures were characterized using scanning electron microscopy (SEM) and X-ray diffraction (XRD). Pellet thermal conductivity was measured by the laser flash method. Significant improvements in thermal conductivity were observed in the as-manufactured pellets with increases of up to 75% at 1073K for the pellets produced from nano-Mo and pre-sintered UO2 granules compared to a pure UO₂ pellet. These results highlight the effectiveness of nano-Mo addition and the pre-sintering of UO2 granules in enhancing the thermal performance of UO₂-based nuclear fuel composites.
{"title":"Fabrication of UO2–Mo composite fuel pellets with enhanced thermal conductivity by using wet mixing","authors":"Fihan Alharbi, Hywel Ragnauth, Timothy Abram, Joel Turner","doi":"10.1016/j.jnucmat.2026.156440","DOIUrl":"10.1016/j.jnucmat.2026.156440","url":null,"abstract":"<div><div>Uranium dioxide (UO₂) is the standard nuclear fuel for light water reactors (LWRs) due to its operational experience, irradiation stability, and ease of manufacture. However, its low thermal conductivity results in a high temperature gradient across the pellet in-service, leading to thermal stresses, deformation, and cracking. This study explores the addition of a high thermal conductivity molybdenum phase as a nano-powder, which is trialled alongside pre-sintered UO<sub>2</sub> granules to improve the thermal conductivity of the resulting pellet. UO<sub>2</sub>- 10 wt. % Mo composite pellets were fabricated by dispersing UO₂ granules and nano-Mo powder in ethanol during mixing, followed by ethanol evaporation and subsequent Spark Plasma Sintering (SPS) at 1473 K with a 5-minute hold. Pellet microstructures were characterized using scanning electron microscopy (SEM) and X-ray diffraction (XRD). Pellet thermal conductivity was measured by the laser flash method. Significant improvements in thermal conductivity were observed in the as-manufactured pellets with increases of up to 75% at 1073K for the pellets produced from nano-Mo and pre-sintered UO<sub>2</sub> granules compared to a pure UO₂ pellet. These results highlight the effectiveness of nano-Mo addition and the pre-sintering of UO<sub>2</sub> granules in enhancing the thermal performance of UO₂-based nuclear fuel composites.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156440"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976100","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 : 2026-03-01Epub Date: 2026-01-02DOI: 10.1016/j.jnucmat.2025.156433
Huan Yao , Changwei Wu , Tianzhou Ye , Junmei Wu , Yingwei Wu , Ping Chen , Qianjin Xia
FeCrAl alloy serves as a promising accident-tolerant fuel cladding material for nuclear reactors. The structural integrity of nuclear fuel throughout the reactor’s service life critically depends on the cladding's creep behavior. Researchers have proposed several constitutive models to predict the high-temperature creep response of FeCrAl alloy tubes. However, the development of reasonable and reliable constitutive models necessitates extensive experimental validation due to the complex interdependencies governing creep behavior. This study aims to establish a new creep constitutive model for FeCrAl alloys through molecular dynamics (MD) simulations and experimental validation. Constitutive model parameters were fitted based on the MD simulation results of high-temperature creep of FeCrAl alloy under a wide range of grain size, temperature, and stress conditions. Subsequently, the key parameters were optimized against the material's bulk mechanical properties and biaxial creep test data. The developed constitutive model was implemented in finite element analysis (FEA) to simulate biaxial creep behavior of thin-walled FeCrAl tubes. A good quantitative agreement was observed between the FEA predictions and the measured results for both axial and hoop creep strain rates. Furthermore, the model is validated against uniaxial creep data from open literature, confirming its reliability in simulating both biaxial and uniaxial creep behavior of FeCrAl cladding tubes. The proposed model can achieve at least an order of magnitude improvement in prediction accuracy over the ORNL (Oak Ridge National Laboratory) model.
{"title":"Creep constitutive model for FeCrAl alloy cladding tube: experiments and molecular dynamics simulations","authors":"Huan Yao , Changwei Wu , Tianzhou Ye , Junmei Wu , Yingwei Wu , Ping Chen , Qianjin Xia","doi":"10.1016/j.jnucmat.2025.156433","DOIUrl":"10.1016/j.jnucmat.2025.156433","url":null,"abstract":"<div><div>FeCrAl alloy serves as a promising accident-tolerant fuel cladding material for nuclear reactors. The structural integrity of nuclear fuel throughout the reactor’s service life critically depends on the cladding's creep behavior. Researchers have proposed several constitutive models to predict the high-temperature creep response of FeCrAl alloy tubes. However, the development of reasonable and reliable constitutive models necessitates extensive experimental validation due to the complex interdependencies governing creep behavior. This study aims to establish a new creep constitutive model for FeCrAl alloys through molecular dynamics (MD) simulations and experimental validation. Constitutive model parameters were fitted based on the MD simulation results of high-temperature creep of FeCrAl alloy under a wide range of grain size, temperature, and stress conditions. Subsequently, the key parameters were optimized against the material's bulk mechanical properties and biaxial creep test data. The developed constitutive model was implemented in finite element analysis (FEA) to simulate biaxial creep behavior of thin-walled FeCrAl tubes. A good quantitative agreement was observed between the FEA predictions and the measured results for both axial and hoop creep strain rates. Furthermore, the model is validated against uniaxial creep data from open literature, confirming its reliability in simulating both biaxial and uniaxial creep behavior of FeCrAl cladding tubes. The proposed model can achieve at least an order of magnitude improvement in prediction accuracy over the ORNL (Oak Ridge National Laboratory) model.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156433"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976105","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 : 2026-03-01Epub Date: 2026-01-01DOI: 10.1016/j.jnucmat.2025.156434
Yingwei Xiong , Xingtong Liu , Wenjuan Wang , Yi Ding
Ceramic immobilization is one of the good strategies for disposal of high-level radioactive waste, but suffers from high temperature and low immobilization capacity. Herein, green spark plasma sintering (SPS) technology was reported to prepare zircon-based multiphase ceramics (x-Z) for enhanced simulated trivalent actinide (Nd3+) immobilization. The effect of Nd3+ content on the phase and microstructure evolutions of the obtained x-Z ceramics was investigated. The x-Z ceramics were prepared by SPS at low sintering temperature (1350 °C) and short time (10 min), and their Nd3+ immobilization capacity was up to 20 at% owing to the low sintering temperature and short sintering time reduced the decomposition of ZrSiO4 phase. In addition, the obtained x-Z ceramics exhibited superior aqueous stability due to the high density achieved by SPS. The green and efficient SPS technology could play a significant role in promoting the industrialization of ceramics immobilization of high-level radioactive waste.
{"title":"Enhanced immobilization of trivalent actinides in zircon-based multiphase ceramics via spark plasma sintering","authors":"Yingwei Xiong , Xingtong Liu , Wenjuan Wang , Yi Ding","doi":"10.1016/j.jnucmat.2025.156434","DOIUrl":"10.1016/j.jnucmat.2025.156434","url":null,"abstract":"<div><div>Ceramic immobilization is one of the good strategies for disposal of high-level radioactive waste, but suffers from high temperature and low immobilization capacity. Herein, green spark plasma sintering (SPS) technology was reported to prepare zircon-based multiphase ceramics (<em>x</em>-Z) for enhanced simulated trivalent actinide (Nd<sup>3+</sup>) immobilization. The effect of Nd<sup>3+</sup> content on the phase and microstructure evolutions of the obtained <em>x</em>-Z ceramics was investigated. The <em>x</em>-Z ceramics were prepared by SPS at low sintering temperature (1350 °C) and short time (10 min), and their Nd<sup>3+</sup> immobilization capacity was up to 20 at% owing to the low sintering temperature and short sintering time reduced the decomposition of ZrSiO<sub>4</sub> phase. In addition, the obtained <em>x</em>-Z ceramics exhibited superior aqueous stability due to the high density achieved by SPS. The green and efficient SPS technology could play a significant role in promoting the industrialization of ceramics immobilization of high-level radioactive waste.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156434"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922720","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}
Beryllium metal is characterized by its unique physical properties, which determines its wide range of applications, including the use in nuclear reactors, resulting inevitably in activated metallic beryllium that has to be treated as radioactive waste. In the present work, the corrosion behavior of metallic beryllium in aqueous NaOH solutions with pH ranging between 6.7 and 14.0 and in solutions simulating the environment in potential waste encapsulation matrices such as Ordinary Portland Cement (OPC) or magnesium phosphate cement (MPC) was studied in detail. Corrosion rates of metallic beryllium samples were experimentally studied by using two direct methods based on gravimetric measurements and the determination of beryllium concentrations in the solution by using Inductively Coupled Plasma - Mass Spectrometry (ICP-MS). A combined method based on these two direct methods is proposed to enable the determination of corrosion rates in various aqueous solutions, including alkaline solutions and those with near neutral pH values. Detailed studies of corroded metal surfaces were carried out using scanning electron microscopy (SEM) combined with energy dispersive X-ray spectroscopy (EDS), indicating pitting corrosion as prominent corrosion mechanism.
{"title":"Corrosion of metallic beryllium in various aqueous solutions","authors":"Andrey Bukaemskiy , Guido Deissmann , Sebastien Caes , Giuseppe Modolo , Dirk Bosbach","doi":"10.1016/j.jnucmat.2026.156465","DOIUrl":"10.1016/j.jnucmat.2026.156465","url":null,"abstract":"<div><div>Beryllium metal is characterized by its unique physical properties, which determines its wide range of applications, including the use in nuclear reactors, resulting inevitably in activated metallic beryllium that has to be treated as radioactive waste. In the present work, the corrosion behavior of metallic beryllium in aqueous NaOH solutions with pH ranging between 6.7 and 14.0 and in solutions simulating the environment in potential waste encapsulation matrices such as Ordinary Portland Cement (OPC) or magnesium phosphate cement (MPC) was studied in detail. Corrosion rates of metallic beryllium samples were experimentally studied by using two direct methods based on gravimetric measurements and the determination of beryllium concentrations in the solution by using Inductively Coupled Plasma - Mass Spectrometry (ICP-MS). A combined method based on these two direct methods is proposed to enable the determination of corrosion rates in various aqueous solutions, including alkaline solutions and those with near neutral pH values. Detailed studies of corroded metal surfaces were carried out using scanning electron microscopy (SEM) combined with energy dispersive X-ray spectroscopy (EDS), indicating pitting corrosion as prominent corrosion mechanism.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156465"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024668","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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