Journal of Nuclear Materials最新文献

{"title":"Point defect segregation at edge dislocations in α-Fe studied by kinetic activation-relaxation technique","authors":"N. Kvashin ,&nbsp;N. Anento ,&nbsp;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":"2026-03-01","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}

{"title":"Effect of salt purity on the corrosion of 316 L SS: Long-term studies in molten FLiNaK and ThF4 - LiF","authors":"Nigel Lucas ,&nbsp;Richard Woods ,&nbsp;Simon Crombleholme ,&nbsp;Hruthik Vandanapu ,&nbsp;Connor Beer ,&nbsp;Jeremie Sobel ,&nbsp;Thomas Steenberg ,&nbsp;Maulik K. Patel","doi":"10.1016/j.jnucmat.2025.156398","DOIUrl":"10.1016/j.jnucmat.2025.156398","url":null,"abstract":"<div><div>This study investigated the influence of impurities on the corrosion behaviour of 316 L stainless steel in molten FLiNaK at 600 °C and LiThF at 700 °C for up to 3000 h. Coupons exposed to untreated FLiNaK containing moisture and oxides exhibited significantly higher mass loss compared to those exposed to purified salt. This coincides with the depletion of chromium (Cr) and iron (Fe) from the steel, as confirmed by ICP-OES analysis of post-experiment salts. SEM analysis identified intergranular corrosion as the primary attack mode in untreated FLiNaK, with corrosion depths up to 112 μm. Conversely, coupons exposed to purified FLiNaK displayed excellent corrosion resistance, despite trace levels of metallic impurities detected in the salt. These impurities are believed to originate from the purification process itself and may contribute to a small degree of observed corrosion. The typical mode of grain boundary Cr dissolution was observed in coupons tested in untreated FLiNaK salts, while this was absent in coupons tested in purified FLiNaK salts. Interestingly, XRD analysis identified chromium carbide phases on coupons from purified FLiNaK tests. Similar trends were observed in LiThF salt, where untreated salt resulted in severe corrosion compared to purified salt. Overall, this study highlights the critical role of salt purity in minimizing corrosion of 316 L stainless steel in molten fluoride salts. The data from the present study indicates that a purified molten FLiNaK at 600 °C can be a long-term non-corrosive environment for 316 L stainless steel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156398"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882272","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":"Homogenization-based irradiation performance analysis of U-Mo/Al dispersion fuel plates","authors":"Wei Li ,&nbsp;Haokun Wang ,&nbsp;Jie Wang ,&nbsp;Qi Lu ,&nbsp;Yuanming Li ,&nbsp;Junmei Wu ,&nbsp;Wenjie Zeng","doi":"10.1016/j.jnucmat.2026.156442","DOIUrl":"10.1016/j.jnucmat.2026.156442","url":null,"abstract":"<div><div>The U-Mo/Al dispersion fuel has emerged as a highly promising candidate for high-density Low Enriched Uranium (LEU) fuel in high-performance research reactors. Understanding the irradiation behavior of U-Mo/Al dispersion fuel plates is crucial for ensuring the safety and performance of these reactors. In this study, we propose a homogenization-based approach to analyze the irradiation performance of U-Mo/Al dispersion fuel plates. At the mesoscopic scale, the Representative Volume Element (RVE) approach, combined with the Finite Element Method (FEM), is employed to determine the equivalent thermomechanical properties of the dispersion fuel meat. To account for the significant influence of the U-Mo/Al interaction layer (IL), the empirical model of the growth of this layer under irradiation is seamlessly integrated into the RVE model. Subsequently, based on the obtained equivalent thermomechanical property models, irradiation performance analysis of U-Mo/Al fuel plate at the macroscopic scale is conducted. Both the meso‑ and macroscale models have been validated against publicly available experimental data. By informing macroscopic performance analysis with mesoscopic homogenization, this study provides a reliable and efficient tool for optimizing the design and enhancing the performance of advanced dispersion fuels in research reactor applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156442"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024748","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":"Processing dictates properties: Mitigating hydrogen embrittlement in U-2Nb alloy by cutting method selection","authors":"Wenliang Xu,&nbsp;Shushan Cui,&nbsp;Shilv Yu,&nbsp;Mengsheng Zhai,&nbsp;Sheng Zhang,&nbsp;Denglei Chen,&nbsp;Yun Fang,&nbsp;Pei Zhang,&nbsp;Dawu Xiao,&nbsp;Tao Fa","doi":"10.1016/j.jnucmat.2026.156459","DOIUrl":"10.1016/j.jnucmat.2026.156459","url":null,"abstract":"<div><div>Tensile specimens of lamellar U-2Nb alloy were prepared using three different cutting methods: fast-speed wire electrical discharge machining (FS-WEDM), slow-speed wire electrical discharge machining (SS-WEDM) and milling. Significant discrepancies in mechanical performance were observed, with the FS-WEDM and SS-WEDM specimens exhibiting total elongations below 10%, which is substantially lower than that achieved by milling. Thermal desorption analysis and fractographic examination indicated that hydrogen trapped at α-U (or UH₃) phases and α/γ_1-2 interfaces is responsible for the hydrogen embrittlement (HE) observed in the WEDM-processed specimens. This conclusion was confirmed by subsequent vacuum dehydrogenation treatments at 393 K, 573 K and 873 K, coupled with corresponding fractographic analysis. Consequently, the utilizing of WEDM should be extremely cautious when preparing tensile specimens for U-2Nb alloy, and milling is recommended as the preferred alternative. This study reveals that WEDM introduces hydrogen, leading to embrittlement of U-2Nb alloy, and provides critical guidance for the sample preparation of HE-sensitive materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156459"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024749","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":"Numerical simulation study on the impact of helium bubbles on hydrogen isotope retention behavior in tungsten","authors":"Wentao Qin ,&nbsp;Weifeng Liu ,&nbsp;Chunjie Niu ,&nbsp;Weiyuan Ni ,&nbsp;Dongping Liu","doi":"10.1016/j.jnucmat.2025.156401","DOIUrl":"10.1016/j.jnucmat.2025.156401","url":null,"abstract":"<div><div>Based on our recent study (Nucl. Fusion 64 (2024) 056,039), an improved model is employed to investigate the impact of helium bubbles in tungsten (W) on hydrogen isotope retention. Simulations were conducted to assess the changes in helium bubble size and density in W under helium pre-irradiation at the W temperature ranging from 400 K to 2300 K. Subsequently, the effect of these helium bubbles on hydrogen isotope retention was examined through simulated hydrogen isotope irradiation at 400 K. As the W surface temperature increases during helium pre-irradiation, the helium bubbles exhibit size growth with concomitant density reduction. Correspondingly, the hydrogen isotope retention increases and then decreases with the rise in W surface temperature during helium pre-irradiation. When the hydrogen dose is further increased, the hydrogen retention gradually increases. The application of rate theory models in our study helps in effectively simulating the long-term evolution and large-scale behavior of hydrogen and helium interactions in W under fusion-relevant conditions, providing a more comprehensive understanding. The research demonstrates that the temperature of helium irradiation significantly influences the characteristics of helium bubbles in W, which in turn affects the retention of hydrogen isotopes. This work contributes to the understanding and optimization of fusion reactor wall materials, promoting improved performance and reliability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156401"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882265","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":"Metastable grain boundary sink behavior revealed through deep-learning image analysis","authors":"Emily Hopkins Mang ,&nbsp;Annie K. Barnett ,&nbsp;Sicong He ,&nbsp;James E. Nathaniel II ,&nbsp;Ryan Jacobs ,&nbsp;Dane Morgan ,&nbsp;Michael Falk ,&nbsp;Jaime Marian ,&nbsp;Mitra L. Taheri","doi":"10.1016/j.jnucmat.2025.156369","DOIUrl":"10.1016/j.jnucmat.2025.156369","url":null,"abstract":"<div><div>Achieving radiation tolerance in crystalline materials requires advancing our understanding of defect evolution and the corresponding grain boundary (GB) response under irradiation. One strategy for realizing more radiation tolerant materials is by tailoring GBs to behave as more efficient defect sinks, however, their non-equilibrium structural response remains insufficiently resolved. In this study, we combine deep-learning enabled object detection used on <em>in situ</em> transmission electron microscopy experiments and multiscale modeling efforts to examine the dynamic relationship between defect microstructure and structural evolution of the GB. To investigate the underlying mechanisms, we employed molecular dynamics simulations, which support the hypothesis that self-healing through point defect (PD) emission modulates the local GB defect concentration that would cumulate as experimentally observable defect density fluctuations. In parallel, experimental observations of time-dependent GB dislocation core evolution and network formation are corroborated by molecular dynamics and physics-based dislocation loop relaxation models suggesting that non-equilibrium GBs can transition to a PD absorption regime modulated by the elastic environment. This work aims to support the idea that GB structural properties are not static, providing unique experimental evidence that non-equilibrium GB structures evolve under extreme conditions and influence the resulting radiation-induced defect microstructure.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156369"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882271","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Expanded method for the determination of burnup in nuclear fuels using multiple neodymium isotopes","authors":"Grant W. Helmreich ,&nbsp;John D. Hunn ,&nbsp;Fred C. Montgomery ,&nbsp;Darren J. Skitt ,&nbsp;Tamara J. Keever ,&nbsp;Benjamin D. Roach ,&nbsp;Kayron T. Rogers ,&nbsp;James W. Sterbenz","doi":"10.1016/j.jnucmat.2025.156411","DOIUrl":"10.1016/j.jnucmat.2025.156411","url":null,"abstract":"<div><div>Established methods for the determination of burnup in nuclear fuels commonly rely on the measurement of ¹⁴⁸Nd in the spent fuel in concert with known fission product yields to determine the atom percent of fissions in the fuel. This isotope of neodymium is used for various reasons, including chemical and radioactive stability, ease of measurement, and low rates of formation and destruction due to neutron flux apart from fission. However, careful calculation of effective cumulative fission yields and correction factors for (n,γ) capture reactions allows for additional stable and long-lived isotopes of neodymium to be used to provide additional independent measurements of burnup, reducing statistical uncertainty. This method was developed and successfully applied to measure the burnup of compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The mean burnup measured using ¹⁴³Nd, ¹⁴⁵Nd, ¹⁴⁶Nd, ¹⁴⁸Nd, and ¹⁵⁰Nd was statistically observed to be the same as that measured using ¹⁴⁸Nd alone, but the statistical uncertainty in the measurement was reduced by a factor of 2, providing a tighter confidence interval in the final results.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156411"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882263","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":"Corrosion study of zirconium alloys via molecular dynamics simulations: I. oxygen migration in zirconium oxides under low-dose irradiation induced displacement damage","authors":"Xiang Li ,&nbsp;Jinsong Zhang ,&nbsp;Shuang Dai ,&nbsp;Ke Wang ,&nbsp;Yi Wang ,&nbsp;Jia Tang ,&nbsp;Shubo Yang ,&nbsp;Qi Cao","doi":"10.1016/j.jnucmat.2025.156419","DOIUrl":"10.1016/j.jnucmat.2025.156419","url":null,"abstract":"<div><div>In zirconium-alloy corrosion models, <em>O²⁻</em> movement through the zirconium oxide film is commonly considered as the rate-determining step. SEM-EDS results in this study indicate that <em>O²⁻</em> movement is driven by the built-in electric field rather than by concentration gradients. Molecular dynamics simulations were employed to investigate <em>O²⁻</em> mobility in several zirconium oxide structures, showing that monoclinic ZrO₂ with a vertical grain boundary provides the highest migration ability. To investigate irradiation effects, the study focuses on low-dose displacement damage resulting from primary knock-on atom events. The simulations including mean square displacement, trajectory and diffusion coefficient results demonstrate that such low dose irradiation induced damage generally causes minimal change in <em>O²⁻</em> mobility, suggesting a limited influence on corrosion under these conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156419"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882264","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":"Small-scale mechanical testing of interfacial toughness in Cr-coated zircaloy-4","authors":"Jiwon Mun ,&nbsp;JungHun Park ,&nbsp;JongDae Hong ,&nbsp;Sebastian Lam ,&nbsp;Peter Hosemann ,&nbsp;Gi-dong Sim ,&nbsp;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":"2026-03-01","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}

{"title":"Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds","authors":"J.T. Rizk,&nbsp;X.-Y. Liu,&nbsp;D.A. Andersson,&nbsp;E. Kardoulaki,&nbsp;N.M. Abdul-Jabbar","doi":"10.1016/j.jnucmat.2025.156409","DOIUrl":"10.1016/j.jnucmat.2025.156409","url":null,"abstract":"<div><div>The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard <em>U</em> model (DFT+<em>U</em>) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+<em>U</em> calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. The DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156409"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922719","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}