Journal of Nuclear Materials最新文献

{"title":"Mechanistic modelling of chemically reactive fission product release during power ramps","authors":"Giacomo Petrosillo ,&nbsp;Jérôme Sercombe ,&nbsp;Laurence Noirot ,&nbsp;Clément Introïni ,&nbsp;Bertrand Baurens ,&nbsp;Lelio Luzzi ,&nbsp;Yves Pontillon","doi":"10.1016/j.jnucmat.2025.156400","DOIUrl":"10.1016/j.jnucmat.2025.156400","url":null,"abstract":"<div><div>This work presents and validates a novel approach for modelling non-inert fission products behaviour during a power transient. The proposed model, MARGARET Active Fission Products, leverages all the microstructural quantities computed by MARGARET-a code originally developed to evaluate the release of inert fission products under normal and incidental loading sequences-to simulate the production, decay, and transport of non-inert fission products. The MARGARET Active Fission Products model is capable of handling 75 different isotopes, ranging from short-lived to long-lived species. The tendency of non-inert fission products to form different chemical compounds in different phases within the fuel compared to inert ones (always gaseous at equilibrium), requires the incorporation of thermochemistry in the calculations, since it directly impacts their release kinetics from the fuel. The thermochemical behaviour of fission products is included by performing calculations with the OpenCalphad thermochemical solver and the TAF-ID database. In this coupled thermochemistry-fission gas release approach, thermochemistry leads to the assessment of the quantity of non-inert fission products in the gas phase, that will percolate towards the free volume and, consequently, be released in the fuel rod.</div><div>To validate the model, a power transient simulation was performed, replicating an experiment conducted on a refabricated fuel rod in the OSIRIS experimental reactor. The simulation results are compared with experimental data, including the total xenon release (measured by puncturing), the release of ¹³⁷Cs, ¹³¹I, and ¹³²Te at each interpellet (obtained from gamma spectrometry), and the radial distribution of xenon in the pellet, before and after the power ramp (analysed by Secondary-Ion Mass Spectrometry measurements). Furthermore, a detailed discussion on the thermochemical results is provided.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156400"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882268","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}

{"title":"Electronic properties and stability of interstitial oxygen in UO2 grain boundaries: An ab initio study","authors":"Ine Arts ,&nbsp;Rolando Saniz ,&nbsp;Gianguido Baldinozzi ,&nbsp;Gregory Leinders ,&nbsp;Marc Verwerft ,&nbsp;Dirk Lamoen","doi":"10.1016/j.jnucmat.2025.156410","DOIUrl":"10.1016/j.jnucmat.2025.156410","url":null,"abstract":"<div><div>The oxidation of UO₂ is primarily governed by the diffusion of oxygen through the lattice. Oxygen diffusion is significantly influenced by defects and interfaces, with grain boundaries being particularly relevant in spent nuclear fuel due to their increasing concentration at the periphery of fuel rods. While experimental studies on bulk uranium suggest defects enhance oxygen diffusion, the role of grain boundaries remains contentious, with discrepancies between theoretical predictions of enhanced diffusion and experimental observations. This study employs density functional theory (DFT+U) to investigate the electronic properties and stability of interstitial oxygen in two coincident site lattice grain boundaries, Σ3 {111} and Σ5 {210}, in UO₂. We compare stoichiometric and non-stoichiometric grain boundary models, examining their formation energies, defect interactions, and local structural distortions. The interstitial oxygen defects cause an expansion of the oxygen cage and a contraction of the U-O bonds, both in bulk and at the grain boundaries. The Σ3 grain boundary showed potential for defect accumulation, while the Σ5 grain boundary did not demonstrate reduced defect formation energies relative to the bulk. Our findings contribute to the understanding the UO₂ oxidation process, in an attempt to address inconsistencies between theoretical and experimental studies on oxygen diffusion in grain boundaries.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156410"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882270","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}

{"title":"Influence of heat treatment on the microstructure and performance of detonation sprayed W/Fe/steel first wall structure","authors":"Canjia Huang ,&nbsp;Zifeng Deng ,&nbsp;Xingli Wang ,&nbsp;Qiang Li ,&nbsp;Wanjing Wang ,&nbsp;Zhilu Liu ,&nbsp;Jieyao He ,&nbsp;Jiaxin Jin ,&nbsp;Wei Cao ,&nbsp;Zongxiao Guo ,&nbsp;Fan Wang ,&nbsp;Yunming Qiu ,&nbsp;Ying Liu ,&nbsp;Chunyan Yu ,&nbsp;Shixing Wang ,&nbsp;Jianjun Huang","doi":"10.1016/j.jnucmat.2025.156414","DOIUrl":"10.1016/j.jnucmat.2025.156414","url":null,"abstract":"<div><div>The fabrication of high-quality W/steel first wall of blanket poses significant challenges in the field of nuclear fusion, primarily due to the substantial differences in thermodynamic properties between tungsten (W) and stainless steel (SS). Detonation sprayed (DS) iron (Fe) coating is a good alternative thermal stress-relieving interlayer between the DS-W armor and the steel substrate. This study adopted two post-heat treatment strategies to optimize the microstructure and improve the performance of the DS-Fe interlayer: carrying out vacuum heat treatments (ranging from 500 °C to 900 °C) on the DS-W/DS-Fe/SS structure and the DS-Fe/SS structure following the deposition of DS-W coating, respectively. The microstructures of the DS-Fe coatings and the DS-W/DS-Fe/SS structures after the heat treatments showed that three sets of these heat-treated specimens had better microstructure improvement than others. The electron beam-based transient high heat flux (HHF) loading tests revealed that the heat-treated specimens demonstrated notable improvements in the performance of the coating structure. Moreover, under transient HHF, the specimens subjected to 900 °C heat treatment after the deposition of DS-W coating exhibited remarkable performance, followed by those treated at 900 °C before the deposition of DS-W coating. The differences in the damage behaviors of the coating structures at 50 MJ/m² HHF between the two sets were studied in order to understand their performance characteristics and distinctions. These results confirmed that holistic heat treatment on the DS-W/DS-Fe/SS structure effectively alleviated the thermal stress in it under the fusion reactor conditions, thereby providing critical references for the manufacturing of the blanket first wall.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156414"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882338","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}

{"title":"Creep constitutive model for FeCrAl alloy cladding tube: experiments and molecular dynamics simulations","authors":"Huan Yao ,&nbsp;Changwei Wu ,&nbsp;Tianzhou Ye ,&nbsp;Junmei Wu ,&nbsp;Yingwei Wu ,&nbsp;Ping Chen ,&nbsp;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}

{"title":"Tensile properties of co-rolled Zr/U-10Mo fuel at ambient and elevated temperatures","authors":"James Zillinger,&nbsp;Jason Schulthess,&nbsp;Randall Scott,&nbsp;Michael Heighes","doi":"10.1016/j.jnucmat.2025.156392","DOIUrl":"10.1016/j.jnucmat.2025.156392","url":null,"abstract":"<div><div>The United States High Performance Research Reactor (USHPRR) initiative is currently interested in replacing high-enriched uranium fuel in research reactors with high-assay, low-enriched uranium fuel. This work aimed to assist in that mission by providing tensile test data on co-rolled Zr/U-10Mo plates at differing thicknesses (0.285 mm, 0.300 mm, and 0.465 mm) and temperatures (293 K, 473 K, and623 K). Yield strength and ultimate tensile strength values are reported along with an in-depth comparison of how temperature, rolling direction, thickness, and the presence of the Zr-interlayer affect mechanical properties. Results showed that yield strength and ultimate tensile strength in Zr/U-10Mo mini-plates are reduced as temperature increases by approximately 300 MPa from 293 to 473 K, and another 100 MPa from 473 to 623 K. Additionally it was found that with increased temperature, there was less differentiation in fuel necking and delamination behavior, as well as a greater perceived elongation to fracture. The elastic modulus of co-rolled miniplates did not follow the volumetric rule of mixtures in terms of averaged strength and is actually lower than the reported moduli of both Zr and U-10Mo.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156392"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922724","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}

{"title":"Corrigendum to “Post-irradiation examination of AGR-3/4 TRISO fuel compacts using three-dimensional X-ray computed tomography” [Journal of Nuclear Materials Volume 620 (2026) 156341]","authors":"Swapnil Morankar ,&nbsp;William C. Chuirazzi ,&nbsp;Rahul R. Kancharla ,&nbsp;Josh J. Kane ,&nbsp;Brian Gross ,&nbsp;Nikolaus L. Cordes ,&nbsp;John D. Stempien","doi":"10.1016/j.jnucmat.2025.156396","DOIUrl":"10.1016/j.jnucmat.2025.156396","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156396"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922740","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}

{"title":"How the presence of molten fluoride salt influences the tellurization of UNS N10003 alloy?","authors":"Rui Wu ,&nbsp;Anbo Ge ,&nbsp;Xin Du ,&nbsp;Yafei Wang ,&nbsp;Shaoqiang Guo","doi":"10.1016/j.jnucmat.2026.156438","DOIUrl":"10.1016/j.jnucmat.2026.156438","url":null,"abstract":"<div><div>Tellurium-induced corrosion in molten salt reactors is strongly influenced by the chemistry of tellurium in molten fluoride salts. In our previous work, Ellingham diagram and <em>E</em>-pTe²⁻ diagrams were constructed to illustrate the theoretical differences between the metal tellurization in tellurium vapor and molten fluoride environments. Here we further validate the electrochemistry of the nickel tellurization reaction in molten fluoride media and highlight the variation of alloy tellurization between the metal tellurization in tellurium vapor and molten fluoride environments. Unlike tellurization in vapor atmosphere, the tellurization of UNS N10003 alloy in molten LiF-NaF-KF eutectic salt shows a strong dependence on electrode potential and only occurs above the Ni/Ni_0.6Te_0.4 reaction potential (1.43 ± 0.03 V). The presence of CrF₃ in the salt influences the tellurium chemistry, thereby accelerating the tellurization of UNS N10003 alloy compared to that observed in the vapor environment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156438"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976104","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}

{"title":"Research progress on residual stress and microcrack control of tungsten fabricated via additive manufacturing","authors":"Jianguo Ma ,&nbsp;Zhihong Liu ,&nbsp;Chunwei Ma ,&nbsp;Wei Wen ,&nbsp;Huapeng Wu ,&nbsp;Haibiao Ji ,&nbsp;Rui Wang ,&nbsp;Yuquan Kuang ,&nbsp;Wangqi Shi ,&nbsp;Haiying Xu ,&nbsp;Weiping Fang ,&nbsp;Zhiyong Wang ,&nbsp;Yetao He","doi":"10.1016/j.jnucmat.2025.156421","DOIUrl":"10.1016/j.jnucmat.2025.156421","url":null,"abstract":"<div><div>This paper systematically investigates the critical challenges in metal additive manufacturing (AM) technology for producing tungsten first wall components in nuclear fusion reactors—namely, residual stress and microcrack control. Research indicates that tungsten's high melting point (3422 °C) combined with AM's rapid cooling characteristics (10³-10⁴ K/s) results in melt pool temperature gradients exceeding 10⁶ K/m, inducing significant residual stresses. while the material exhibits a sharp decline in plasticity within the brittle-to-ductile transition temperature range of 450–650 K, making residual tensile stresses (&gt;100 MPa) prone to triggering microcracks. Synergistic regulation through optimized scanning strategies (e.g., adjusting scan paths and energy distribution) and post-processing techniques achieves the following: Surface strengthening techniques (e.g., laser shock peening) form a 100–300 μm compressive stress layer, significantly reducing microcrack density; High-temperature heat treatment (1200–1400 °C) promotes crack healing through grain boundary migration. Data-driven machine learning methods enable real-time residual stress prediction, supporting intelligent process optimization. Future efforts should integrate multiscale simulations with irradiation experiments to validate and advance the engineering application of tungsten components in fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156421"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922654","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}

{"title":"Sustainable granite encapsulation of volatile nuclear wastes: A flux-mediated low-temperature strategy","authors":"Wenhong Han ,&nbsp;Xiaoyan Shu ,&nbsp;Wencai Cheng ,&nbsp;Guilin Wei ,&nbsp;Yuxuan He ,&nbsp;Ran Tan ,&nbsp;Mingfen Wen ,&nbsp;Xirui Lu","doi":"10.1016/j.jnucmat.2026.156441","DOIUrl":"10.1016/j.jnucmat.2026.156441","url":null,"abstract":"<div><div>To minimize the loss of complex low-melting-point volatile nuclear waste (primarily containing Cs, Sr, Ba, etc.) during the solidification process, B₂O₃ was employed as an auxiliary agent to achieve low-temperature immobilization in amphibolite granulite. The phase evolution, microstructure, elemental distribution, and glass-to-crystal transition of the resulting matrices were systematically characterized. Furthermore, the underlying mechanism by which structural changes of element B in the solidified form affects its properties was explored through density functional theory (DFT) calculations. The results demonstrate that the incorporation of B₂O₃ effectively reduces the sintering temperature by 300°C (from 1500°C to 1200°C). The obtained solidified form exhibits excellent durability, with normalized leaching rates of Cs, Sr, and Ba reaching as low as ∼10⁻³ g·m⁻²·d⁻¹ after 42 days. The formation of a CsAlSiO₄ crystalline phase contributes to retarding the leaching kinetics of radionuclides, thus enhancing the chemical durability. In summary, utilizing B₂O₃ as an auxiliary agent enables amphibolite granulite to leverage the synergistic advantages of glass and ceramic phases, effectively immobilizing volatile TRPO-CS waste (with retention rates of approximately 95% for Cs and 99% for Sr). This strategy provides a promising pathway for the safe and green disposal of radioactive waste.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156441"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922721","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}

{"title":"Origin of the inner-layer stratification of 316 L in 430℃ high-temperature steam","authors":"Haodong Wu ,&nbsp;Yaqing Ren ,&nbsp;Xiangguo Li ,&nbsp;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 (&lt;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_i) and O. DFT calculations showed that the diffusion barrier of H_i in FeOOH was higher than that in Fe₂O₃ and Cr₂O₃. In a low-DO environment (&lt;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":"2026-03-01","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}