Pub 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":"2026-01-05","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 : 2026-01-03DOI: 10.1016/j.jnucmat.2026.156437
Haodong Wu , Yaqing Ren , Xiangguo Li , Jian Xu
The oxide stratification behavior of the oxide on 316 L stainless steel (SS) in high-temperature steam at 430 °C with different dissolved oxygen (DO) concentrations (<5/50/200 ppb) was systematically studied. The oxide layer has a double-layer structure. However, the stratification of the inner layer is related to DO. In the high DO environment (50/200 ppb), the Cr-Fe oxide layer exhibited a certain element stratification, owing to the combined action of interstitial hydrogen (Hi) and O. DFT calculations showed that the diffusion barrier of Hi in FeOOH was higher than that in Fe2O3 and Cr2O3. In a low-DO environment (<5 ppb), the transition from internal to external oxidation is driven by the lack of oxygen. The Cr content in the Cr-rich layer was significantly higher than that in the matrix. The distribution of Ni in the Ni-enriched layer was related to that of the Cr-rich layer.
{"title":"Origin of the inner-layer stratification of 316 L in 430℃ high-temperature steam","authors":"Haodong Wu , Yaqing Ren , Xiangguo Li , Jian Xu","doi":"10.1016/j.jnucmat.2026.156437","DOIUrl":"10.1016/j.jnucmat.2026.156437","url":null,"abstract":"<div><div>The oxide stratification behavior of the oxide on 316 L stainless steel (SS) in high-temperature steam at 430 °C with different dissolved oxygen (DO) concentrations (<5/50/200 ppb) was systematically studied. The oxide layer has a double-layer structure. However, the stratification of the inner layer is related to DO. In the high DO environment (50/200 ppb), the Cr-Fe oxide layer exhibited a certain element stratification, owing to the combined action of interstitial hydrogen (H<sub>i</sub>) and O. DFT calculations showed that the diffusion barrier of H<sub>i</sub> in FeOOH was higher than that in Fe<sub>2</sub>O<sub>3</sub> and Cr<sub>2</sub>O<sub>3</sub>. In a low-DO environment (<5 ppb), the transition from internal to external oxidation is driven by the lack of oxygen. The Cr content in the Cr-rich layer was significantly higher than that in the matrix. The distribution of Ni in the Ni-enriched layer was related to that of the Cr-rich layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156437"},"PeriodicalIF":3.2,"publicationDate":"2026-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922723","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 : 2026-01-02DOI: 10.1016/j.jnucmat.2026.156435
Zhe Liu , Zhihao Wang , Ding Zuo , Wenbo Liu , Huiqun Liu , Ruiqian Zhang
The microstructure and high temperature mechanical properties of annealed FeCrAl-Mo/Nb alloy were studied in this paper. The effects of alloying elements Mo and Nb on the microstructure, recrystallization behavior and mechanical properties of FeCrAl alloy were systematically analyzed, and the mechanism was discussed. The results show that the recrystallization behavior of FeCrAl alloy with high Mo content is significantly delayed because more solid solution atoms hinder the dislocation movement. The recrystallization behavior of FeCrAl-2Mo0.65Nb alloy is promoted by the particles stimulated recrystallization nucleation due to the existence of Laves phase at the initial stage of recrystallization. At the later stage of recrystallization, the recrystallized grains of this alloy are not easy to grow due to the pinning effect of Laves phase on the grain boundary, and the average grain size of fully recrystallized grains is small, which is 6.51 μm. The main recrystallization mechanism of FeCrAl alloy is strain induced grain boundary migration nucleation and growth. The addition of Mo and Nb has no obvious effect on the recrystallization texture type of FeCrAl alloy, but it would form different maximum texture strength. And the room temperature hardness of the alloy is improved by solution strengthening and second phase strengthening, respectively. The contribution of Laves phase to the high temperature strength of FeCrAl alloy is limited. The FeCrAl-2Mo0.65Nb alloy with partially recrystallized microstructure shows relatively good strength and ductility at 600 °C. Due to the existence of high-density dislocations and Laves phase, the alloy has relatively large displacement and deceleration creep time at 400 °C.
研究了退火后的feral - mo /Nb合金的显微组织和高温力学性能。系统分析了合金元素Mo和Nb对FeCrAl合金组织、再结晶行为和力学性能的影响，并探讨了其作用机理。结果表明，高Mo含量的FeCrAl合金的再结晶行为明显延迟，因为更多的固溶体原子阻碍了位错的移动。FeCrAl-2Mo0.65Nb合金的再结晶行为是由再结晶初期Laves相的存在引起的颗粒激发的再结晶成核促进的。在再结晶后期，由于Laves相在晶界上的钉钉作用，合金的再结晶晶粒不易长大，完全再结晶晶粒的平均晶粒尺寸较小，为6.51 μm。FeCrAl合金的再结晶机制主要是应变诱导晶界迁移、形核和长大。Mo和Nb的加入对FeCrAl合金的再结晶织构类型没有明显影响，但会形成不同的最大织构强度。通过固溶强化和第二相强化分别提高了合金的室温硬度。Laves相对FeCrAl合金高温强度的贡献是有限的。部分再结晶组织的feral - 2mo0.65 nb合金在600℃时表现出较好的强度和塑性。由于高密度位错和Laves相的存在，合金在400℃时具有较大的位移和减速蠕变时间。
{"title":"Effect of Mo/Nb addition on recrystallization behavior and high temperature mechanical properties of FeCrAl alloy tubes","authors":"Zhe Liu , Zhihao Wang , Ding Zuo , Wenbo Liu , Huiqun Liu , Ruiqian Zhang","doi":"10.1016/j.jnucmat.2026.156435","DOIUrl":"10.1016/j.jnucmat.2026.156435","url":null,"abstract":"<div><div>The microstructure and high temperature mechanical properties of annealed FeCrAl-Mo/Nb alloy were studied in this paper. The effects of alloying elements Mo and Nb on the microstructure, recrystallization behavior and mechanical properties of FeCrAl alloy were systematically analyzed, and the mechanism was discussed. The results show that the recrystallization behavior of FeCrAl alloy with high Mo content is significantly delayed because more solid solution atoms hinder the dislocation movement. The recrystallization behavior of FeCrAl-2Mo0.65Nb alloy is promoted by the particles stimulated recrystallization nucleation due to the existence of Laves phase at the initial stage of recrystallization. At the later stage of recrystallization, the recrystallized grains of this alloy are not easy to grow due to the pinning effect of Laves phase on the grain boundary, and the average grain size of fully recrystallized grains is small, which is 6.51 μm. The main recrystallization mechanism of FeCrAl alloy is strain induced grain boundary migration nucleation and growth. The addition of Mo and Nb has no obvious effect on the recrystallization texture type of FeCrAl alloy, but it would form different maximum texture strength. And the room temperature hardness of the alloy is improved by solution strengthening and second phase strengthening, respectively. The contribution of Laves phase to the high temperature strength of FeCrAl alloy is limited. The FeCrAl-2Mo0.65Nb alloy with partially recrystallized microstructure shows relatively good strength and ductility at 600 °C. Due to the existence of high-density dislocations and Laves phase, the alloy has relatively large displacement and deceleration creep time at 400 °C.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156435"},"PeriodicalIF":3.2,"publicationDate":"2026-01-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922718","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 : 2026-01-02DOI: 10.1016/j.jnucmat.2026.156436
N. Kvashin , N. Anento , L. Malerba
The mechanical properties of crystalline materials such as metals, are strongly related to the mobility of dislocations, which is directly affected by their interaction with other defects present in the microstructure and acting as obstacles. Under irradiation conditions the number density of point defects increases substantially, leading to several phenomena at the atomic scale, some of which are related with the behaviour of dislocations as sinks for vacancies and self-interstitial atoms. In this work we present an in-depth study of the segregation process of point defects to an edge dislocation in α-Fe, performed with an on-the-fly kinetic Monte Carlo model, the kinetic activation-relaxation technique (k-ART). Our KMC simulations show that, in the vicinity of the dislocation core, the dynamics of vacancies and SIAs is accelerated before absorption. For the former, the preferential path is along the compression region while for the latter is along the tensile region. This work therefore provides a greater knowledge of the dynamic properties of point defects around of dislocations, such as free migration time, acceleration/deceleration of point defects motion and energies of absorption events. These results will allow more precise modelling of the microstructure evolution of polycrystalline materials, improving the predictive capabilities of existing models in the long term. In order to ensure transferability of these findings to other KMC models, the data obtained in the simulations have been used to train a prediction model based on a Machine Learning logistic regression algorithm.
{"title":"Point defect segregation at edge dislocations in α-Fe studied by kinetic activation-relaxation technique","authors":"N. Kvashin , N. Anento , L. Malerba","doi":"10.1016/j.jnucmat.2026.156436","DOIUrl":"10.1016/j.jnucmat.2026.156436","url":null,"abstract":"<div><div>The mechanical properties of crystalline materials such as metals, are strongly related to the mobility of dislocations, which is directly affected by their interaction with other defects present in the microstructure and acting as obstacles. Under irradiation conditions the number density of point defects increases substantially, leading to several phenomena at the atomic scale, some of which are related with the behaviour of dislocations as sinks for vacancies and self-interstitial atoms. In this work we present an in-depth study of the segregation process of point defects to an edge dislocation in α-Fe, performed with an on-the-fly kinetic Monte Carlo model, the kinetic activation-relaxation technique (k-ART). Our KMC simulations show that, in the vicinity of the dislocation core, the dynamics of vacancies and SIAs is accelerated before absorption. For the former, the preferential path is along the compression region while for the latter is along the tensile region. This work therefore provides a greater knowledge of the dynamic properties of point defects around of dislocations, such as free migration time, acceleration/deceleration of point defects motion and energies of absorption events. These results will allow more precise modelling of the microstructure evolution of polycrystalline materials, improving the predictive capabilities of existing models in the long term. In order to ensure transferability of these findings to other KMC models, the data obtained in the simulations have been used to train a prediction model based on a Machine Learning logistic regression algorithm.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156436"},"PeriodicalIF":3.2,"publicationDate":"2026-01-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922742","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 : 2026-01-02DOI: 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":"2026-01-02","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 : 2026-01-01DOI: 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":"2026-01-01","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}
Pub Date : 2025-12-30DOI: 10.1016/j.jnucmat.2025.156424
Yu Wang , Heng Chen , Rui Su , Bin Xu , Rulong Zhou , Dongdong Li , Yu-Wei You , Pengfei Guan , Changsong Liu
This work develops a high-accuracy artificial neural network (ANN) potential for osmium (Os) to enable large-scale irradiation damage simulations in fusion materials. The potential employs spherical harmonic-Chebyshev polynomial descriptors within a Behler-Parrinello neural network architecture, trained on an extensive dataset generated via density functional theory and ab initio molecular dynamics. Comprehensive validations demonstrate excellent agreement with reference calculations and experimental data across multiple properties: lattice constants of diverse crystal structures, elastic constants for hexagonal close-packed Os, dimer interactions, and defect formation energies (vacancies, interstitials, surfaces). The ANN potential accurately reproduces key behaviors under extreme conditions, including melting characteristics, sputtering thresholds, and primary knock-on atom collision cascades. Simulations reveal defect evolution and clustering during radiation events. This transferable potential provides a critical computational tool for investigating Os precipitation effects on tritium retention and irradiation hardening in tungsten-based plasma-facing materials for fusion reactors.
{"title":"Development of a neural network potential for osmium enables irradiation damage simulations","authors":"Yu Wang , Heng Chen , Rui Su , Bin Xu , Rulong Zhou , Dongdong Li , Yu-Wei You , Pengfei Guan , Changsong Liu","doi":"10.1016/j.jnucmat.2025.156424","DOIUrl":"10.1016/j.jnucmat.2025.156424","url":null,"abstract":"<div><div>This work develops a high-accuracy artificial neural network (ANN) potential for osmium (Os) to enable large-scale irradiation damage simulations in fusion materials. The potential employs spherical harmonic-Chebyshev polynomial descriptors within a Behler-Parrinello neural network architecture, trained on an extensive dataset generated via density functional theory and <em>ab initio</em> molecular dynamics. Comprehensive validations demonstrate excellent agreement with reference calculations and experimental data across multiple properties: lattice constants of diverse crystal structures, elastic constants for hexagonal close-packed Os, dimer interactions, and defect formation energies (vacancies, interstitials, surfaces). The ANN potential accurately reproduces key behaviors under extreme conditions, including melting characteristics, sputtering thresholds, and primary knock-on atom collision cascades. Simulations reveal defect evolution and clustering during radiation events. This transferable potential provides a critical computational tool for investigating Os precipitation effects on tritium retention and irradiation hardening in tungsten-based plasma-facing materials for fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156424"},"PeriodicalIF":3.2,"publicationDate":"2025-12-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922741","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 : 2025-12-30DOI: 10.1016/j.jnucmat.2025.156426
Jiwon Mun , JungHun Park , JongDae Hong , Sebastian Lam , Peter Hosemann , Gi-dong Sim , Ho Jin Ryu
This study quantifies fracture mechanisms and interfacial fracture toughness in Cr-coated Zircaloy-4 tube systems proposed as accident-tolerant fuel (ATF) cladding, using deep-notched (DN) microtensile specimens. Three gauge configurations were tested: single-phase Zr (DN-Zr), single-phase Cr (DN-Cr), and a Zr/Cr interface located at the notch root (DN-interface). The mode I stress intensity factor (SIF) for the ideal sharp-notch geometry was obtained from an analytical correlation and calibrated by finite-element (FE) J-integral analysis (Contour Integral method using ABAQUS/Standard), showing close agreement and validating the modeling. Using this calibration, the interface fracture toughness for the DN-interface configuration was extracted. Interfacial failure exhibits two distinct modes: specimens exhibiting interface-crossing yield = 1.89 ± 0.20 MPa, while delamination-dominated specimens yield = 1.12 ± 0.02 MPa. For safety assessments of Cr-coated Zircaloy-4 applications, we suggest a conservative, weakest-mode design input of = 1.12 MPa. The calibrated methodology, combined with the conservative interfacial toughness , enables quantitative screening and optimization of coating–interface configurations for ATF cladding.
{"title":"Small-scale mechanical testing of interfacial toughness in Cr-coated zircaloy-4","authors":"Jiwon Mun , JungHun Park , JongDae Hong , Sebastian Lam , Peter Hosemann , Gi-dong Sim , Ho Jin Ryu","doi":"10.1016/j.jnucmat.2025.156426","DOIUrl":"10.1016/j.jnucmat.2025.156426","url":null,"abstract":"<div><div>This study quantifies fracture mechanisms and interfacial fracture toughness in Cr-coated Zircaloy-4 tube systems proposed as accident-tolerant fuel (ATF) cladding, using deep-notched (DN) microtensile specimens. Three gauge configurations were tested: single-phase Zr (DN-Zr), single-phase Cr (DN-Cr), and a Zr/Cr interface located at the notch root (DN-interface). The mode I stress intensity factor (SIF) <span><math><msub><mi>K</mi><mi>I</mi></msub></math></span> for the ideal sharp-notch geometry was obtained from an analytical correlation and calibrated by finite-element (FE) J-integral analysis (Contour Integral method using ABAQUS/Standard), showing close agreement and validating the modeling. Using this calibration, the interface fracture toughness for the DN-interface configuration <span><math><msub><mi>K</mi><mrow><mi>Q</mi><mo>,</mo><mtext>int</mtext></mrow></msub></math></span> was extracted. Interfacial failure exhibits two distinct modes: specimens exhibiting interface-crossing yield <span><math><msub><mi>K</mi><mrow><mi>Q</mi><mo>,</mo><mtext>int</mtext></mrow></msub></math></span> = 1.89 ± 0.20 MPa<span><math><mrow><mi>√</mi><mi>m</mi></mrow></math></span>, while delamination-dominated specimens yield <span><math><msub><mi>K</mi><mrow><mi>Q</mi><mo>,</mo><mtext>int</mtext></mrow></msub></math></span> = 1.12 ± 0.02 MPa<span><math><mrow><mi>√</mi><mi>m</mi></mrow></math></span>. For safety assessments of Cr-coated Zircaloy-4 applications, we suggest a conservative, weakest-mode design input of <span><math><msubsup><mi>K</mi><mrow><mi>Q</mi><mo>,</mo><mtext>int</mtext></mrow><mi>cons</mi></msubsup></math></span> = 1.12 MPa<span><math><mrow><mi>√</mi><mi>m</mi></mrow></math></span>. The calibrated methodology, combined with the conservative interfacial toughness <span><math><msubsup><mi>K</mi><mrow><mi>Q</mi><mo>,</mo><mtext>int</mtext></mrow><mi>cons</mi></msubsup></math></span>, enables quantitative screening and optimization of coating–interface configurations for ATF cladding.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156426"},"PeriodicalIF":3.2,"publicationDate":"2025-12-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922717","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 : 2025-12-30DOI: 10.1016/j.jnucmat.2025.156425
Alisha J. Cramer , Peter G. Martin , Thomas B. Scott
Depleted uranium materials pose a significant challenge with respect to their safe, long-term storage, not only from a radiological standpoint, but also with regards to fire safety owing to the reactivity of finely-divided uranium material with air. The risk of pyrophoric ignition is known to be strongly dependent on the material’s specific surface area, however, other factors are known to likely play a role in either increasing or decreasing such a risk. In this article, the research to date on the pyrophoric behaviour of uranium compounds is compiled and examined with a view to understanding the factors controlling pyrophoric ignition and determining the gaps in current knowledge. Although some influencing factors, such as specific surface area, dominate ignition behaviour, several other factors can disrupt expected behaviour, demonstrating that pyrophoric ignition is a complex, dynamic interplay of different factors.
{"title":"Pyrophoricity of uranium and uranium compounds: Mechanisms, knowledge gaps, and implications for nuclear safety","authors":"Alisha J. Cramer , Peter G. Martin , Thomas B. Scott","doi":"10.1016/j.jnucmat.2025.156425","DOIUrl":"10.1016/j.jnucmat.2025.156425","url":null,"abstract":"<div><div>Depleted uranium materials pose a significant challenge with respect to their safe, long-term storage, not only from a radiological standpoint, but also with regards to fire safety owing to the reactivity of finely-divided uranium material with air. The risk of pyrophoric ignition is known to be strongly dependent on the material’s specific surface area, however, other factors are known to likely play a role in either increasing or decreasing such a risk. In this article, the research to date on the pyrophoric behaviour of uranium compounds is compiled and examined with a view to understanding the factors controlling pyrophoric ignition and determining the gaps in current knowledge. Although some influencing factors, such as specific surface area, dominate ignition behaviour, several other factors can disrupt expected behaviour, demonstrating that pyrophoric ignition is a complex, dynamic interplay of different factors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156425"},"PeriodicalIF":3.2,"publicationDate":"2025-12-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882267","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}
Copper (Cu)-based materials suffer from performance degradation due to the accumulation of radiation-induced defects, limiting their application in high-radiation environments. The van der Waals (vdW) type interface formed by introducing graphene (Gr) into Cu-based materials is expected to address the challenge of modifying dislocation characteristics in metallic interface engineering. However, specific methods to modify the interface and their impacts on radiation resistance still need further investigation and quantification. Here, we investigate the interfacial vdW dislocation in Cu/Gr composites, in order to enhance radiation resistance via atomic simulations. The results reveal that increasing Gr rotation angle can lead to a rise in the interfacial dislocation density, while increasing Gr thickness broadens dislocation core width. Quantitative analysis reveals the optimal structural parameters: a critical 15° rotation angle and 4-layer thickness of Gr correspond to point defects reductions of up to 19.5 % and 35.6 %, respectively, compared to the unmodified interface. These findings are crucial for understanding and designing new vdW/metal composites with enhanced radiation resistance.
{"title":"Interfacial dislocation engineering in copper-graphene composites: Atomic insights into enhanced radiation resistance","authors":"Qi Zhang, Zhuoxin Yan, Zhe Yan, Boan Zhong, Mingyu Gong, Yue Liu, Tongxiang Fan","doi":"10.1016/j.jnucmat.2025.156423","DOIUrl":"10.1016/j.jnucmat.2025.156423","url":null,"abstract":"<div><div>Copper (Cu)-based materials suffer from performance degradation due to the accumulation of radiation-induced defects, limiting their application in high-radiation environments. The van der Waals (vdW) type interface formed by introducing graphene (Gr) into Cu-based materials is expected to address the challenge of modifying dislocation characteristics in metallic interface engineering. However, specific methods to modify the interface and their impacts on radiation resistance still need further investigation and quantification. Here, we investigate the interfacial vdW dislocation in Cu/Gr composites, in order to enhance radiation resistance via atomic simulations. The results reveal that increasing Gr rotation angle can lead to a rise in the interfacial dislocation density, while increasing Gr thickness broadens dislocation core width. Quantitative analysis reveals the optimal structural parameters: a critical 15° rotation angle and 4-layer thickness of Gr correspond to point defects reductions of up to 19.5 % and 35.6 %, respectively, compared to the unmodified interface. These findings are crucial for understanding and designing new vdW/metal composites with enhanced radiation resistance.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156423"},"PeriodicalIF":3.2,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976099","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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