Journal of Nuclear Materials最新文献

{"title":"Interactions of noble gas atoms with vacancies, interstitials and alloying titanium/chromium in vanadium","authors":"Mingliang Wei ,&nbsp;Pengbo Zhang ,&nbsp;Shuxian Sun ,&nbsp;Guiqiu Wang ,&nbsp;Yichao Wang ,&nbsp;Yaxia Wei ,&nbsp;Pengfei Zheng","doi":"10.1016/j.jnucmat.2025.155805","DOIUrl":"10.1016/j.jnucmat.2025.155805","url":null,"abstract":"<div><div>The presence of noble gas atoms has a large effect on the properties of nuclear materials, such as hardening and embrittlement. In this study, the clustering behavior of neon (Ne), argon (Ar) and krypton (Kr) atoms in vanadium and their interaction with vacancies/interstitials and alloying titanium/chromium (Ti/Cr) were investigated through first-principles calculations. Ne-Ne pairs exhibit attractive interactions with the binding energy of 0.75 eV, while most Ar-Ar and Kr-Kr pairs show repulsive interactions with the binding energies of −0.85 eV and −0.62 eV, respectively. Ne atoms tend to form interstitial clusters via the self-trapping effect and pre-existing vacancies further enhancing the cluster stability, whereas Ar and Kr clusters are less stable. Furthermore, the interactions of noble gas atoms with vacancy and self-interstitial atoms (SIAs) were determined to explore the effect of Ne on point defect behaviors. Ne can drive SIA migration and promote the recombination of SIAs and vacancies, while Ar and Kr primarily promote the diffusion and evolution of SIAs. Additionally, the alloying Ti can attract three noble gas atoms, thereby promoting their solution and retention, whereas Cr repels them. Finally, the effects of temperature and Ti concentration on the effective diffusivity of Ne, Ar and Kr were predicted using empirical formulas. This work deepens the understanding about the behavior of noble gas atoms and the interaction of point defects in vanadium alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155805"},"PeriodicalIF":2.8,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143814980","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":"Heterogeneous nucleation of plastic defects and tension-compression asymmetry in the presence of vacancies in W single crystals","authors":"Ziyi Li,&nbsp;Wensheng Liu,&nbsp;Yunzhu Ma,&nbsp;Chaoping Liang","doi":"10.1016/j.jnucmat.2025.155806","DOIUrl":"10.1016/j.jnucmat.2025.155806","url":null,"abstract":"<div><div>The tension–compression asymmetry with pre-existing vacancies is investigated for tungsten using molecular dynamics (MD). The tension–compression asymmetry is revealed by means of uniaxial tension and compression along [100], [110], [111], and [112] crystallographic orientations with different strain rates (10⁸ ∼ 10¹¹ s^-1). Results show that except for [110] loading orientation, the yield stresses in compressive are generally greater than those in tensile loading. Vacancy narrows the tension–compression asymmetry as it reduces the gap between tensile and compressive yield strengths when the vacancy concentration goes up. This is through the coalescence of individual vacancy into vacancy clusters before yielding. Aggregation and coalescence of vacancies before yielding lead to the formation of different types of defects, facilitating plastic deformations at yielding. Thus, various plastic deformation mechanisms, like vacancy dislocation loops, twinning, anti-twinning, etc., are observed in tension and compression along different crystallographic orientations. Owing to the non-planar cores, <span><math><mrow><mn>1</mn><mo>/</mo><mn>2</mn><mo>〈</mo><mn>111</mn><mo>〉</mo></mrow></math></span> screw dislocation is identified as the manipulator behind those plastic deformations. The critical resolved shear stress (CRSSs) on the maximum resolved shear stress plane (MRSSP) for the <span><math><mrow><mn>1</mn><mo>/</mo><mn>2</mn><mo>[</mo><mn>111</mn><mo>]</mo></mrow></math></span> screw dislocation loaded in tension and compression are determined and responsible for the origin of tension–compression asymmetry in tungsten.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155806"},"PeriodicalIF":2.8,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143807729","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":"High-temperature nanoindentation creep studies on castable and sintered nanostructured low-activation ferritic-martensitic alloys","authors":"A. Sharma ,&nbsp;M. Ouyang ,&nbsp;E.D. Hintsala ,&nbsp;D. Stauffer ,&nbsp;W. Zhong ,&nbsp;Y. Yang ,&nbsp;J.R. Trelewicz ,&nbsp;L.L. Snead ,&nbsp;D.J. Sprouster","doi":"10.1016/j.jnucmat.2025.155804","DOIUrl":"10.1016/j.jnucmat.2025.155804","url":null,"abstract":"<div><div>In this article, we present the creep characteristics of two reduced activation ferritic-martensitic steels of identical starting compositions formed by different fabrication routes: a nanostructured ferritic alloy commonly referred to as a castable nanostructured alloy (CNA) and a sintered nanostructured alloy (SNA) variant. Through a series of nanoindentation experiments spanning a temperature range of 25 °C to 650 °C, with a maximum load of 100 mN, we find creep behaviors in the cast and sintered materials to be remarkably similar. The creep stress exponent (<span><math><mi>n</mi></math></span>) for CNA and SNA were found to be in the range of 8–35 and the activation volume was ∼14–42<span><math><msup><mrow><mi>b</mi></mrow><mn>3</mn></msup></math></span>, underscoring a dominance of dislocation-mediated mechanisms in both alloys. Notably, we observed a decline in the creep stress exponent with increasing temperature, attributable to the heightened influence of thermally activated dislocations. This phenomenon suggests a potential transition in the deformation mechanism towards a thermally activated dislocation climb process, significantly impacting the observed creep behavior.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"611 ","pages":"Article 155804"},"PeriodicalIF":2.8,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143815358","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":"The failure mechanism of nuclear-grade 316H protective oxide layer in long-term supercritical CO2 Corrosion","authors":"Zhihao Wang ,&nbsp;Jichun Zou ,&nbsp;Minghao Wang ,&nbsp;Shen Li ,&nbsp;Dequan Peng ,&nbsp;Shuai Chen ,&nbsp;Wanhuan Yang ,&nbsp;Weihua Zhong ,&nbsp;Wen Yang","doi":"10.1016/j.jnucmat.2025.155803","DOIUrl":"10.1016/j.jnucmat.2025.155803","url":null,"abstract":"<div><div>An experimental investigation was conducted to evaluate the corrosion behavior of nuclear-grade 316H austenitic stainless steel in supercritical CO₂ (S-CO₂) at 500 °C and 25 MPa over 6400 h. The analysis of the results indicated that, during the initial stage of corrosion, a protective Cr₂O₃ layer formed on the material's surface, exhibiting significant corrosion resistance. However, a sharp escalation in weight gain (13.4 × increase) was observed after 3200 h, accompanied by the formation of a dual-layered oxide structure (Fe₃O₄/FeCr₂O₄), indicating the failure of the Cr₂O₃ layer. This study elucidates the intrinsic chemical failure (InCF) mechanism of Cr₂O₃ under prolonged S-CO₂ exposure, providing critical insights for material selection in advanced nuclear systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"611 ","pages":"Article 155803"},"PeriodicalIF":2.8,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143821048","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":"A critical analysis of U-Pu-Zr phase transitions using calorimetric, microstructural, and phase equilibria data","authors":"Scott Middlemas,&nbsp;Cynthia Adkins","doi":"10.1016/j.jnucmat.2025.155778","DOIUrl":"10.1016/j.jnucmat.2025.155778","url":null,"abstract":"<div><div>Metallic fuels consisting primarily of uranium, plutonium, and zirconium (U-Pu-Zr) are a leading material candidate for fast-spectrum nuclear reactors. Early demonstration programs proved the principle of safe and efficient fast reactor operation, however there is still considerable uncertainty regarding the phase equilibria and microstructural evolution across the ternary composition space. Quantitative phase formation and identification measurements are scarce and often incomplete, with studies reporting either phase transition temperatures or phase identification data, but not both from the same specimens. In this study, we critically compared experimental and calculated phase transition data and correlated with the microstructure and phase characterization data of as-cast and annealed U-Pu-Zr alloys. Differential scanning calorimetry (DSC) was used to measure phase transitions in the subsolidus regions (723−948 K) of three ternary U-Pu-Zr alloys with similar plutonium concentrations but various U/Zr ratios. Due to sluggish kinetics and narrow ranges of phase stability, complex peaks required the use of a Frazier-Suzuki peak fitting algorithm to deconvolute and calculate transition peak temperatures and enthalpies. We also identified trends of phase transition behavior by critically comparing our DSC data with previous phase transition measurements as well as historical and calculated phase equilibrium diagrams. This provides a critical approach for benchmarking and assessing the quality of new U-Pu-Zr phase equilibria data prior to its incorporation into nuclear material databases.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155778"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143837761","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":"Low-temperature sintering of (U,Pu)O2 MOX in mild oxidative conditions","authors":"Jacobus Boshoven ,&nbsp;Jean-François Vigier ,&nbsp;Philipp Pöml ,&nbsp;Abibatou Ndiaye ,&nbsp;Bertrand Morel ,&nbsp;Rudy J.M. Konings ,&nbsp;Karin Popa ,&nbsp;Marco Cologna","doi":"10.1016/j.jnucmat.2025.155800","DOIUrl":"10.1016/j.jnucmat.2025.155800","url":null,"abstract":"<div><div>We compare typical reductive sintering conditions for U, Pu mixed oxides (4 h at 1700°C in Ar/6% H₂ + 1200 ppm H₂O) with lower temperature and mildly oxidative conditions (2 h at 1200°C in CO/CO₂ = 1/9) and report on the resulting microstructures and homogeneity. We show that lower temperature and mildly oxidative conditions, without cover gas change, can give close-to stoichiometric, crack-free MOX pellets with a relative density of ∼ 95%, and we propose ways to improve the homogenisations of PuO₂ and UO₂ and increase the grain size.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155800"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783935","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development of the Technological Process for the IGR Reactor's Highly-Enriched Irradiated Uranium-Graphite Fuel Immobilization","authors":"Kuanysh K. Samarkhanov ,&nbsp;Yuliya Yu. Baklanova ,&nbsp;Olga S. Bukina ,&nbsp;Viktor V. Baklanov ,&nbsp;Yerbolat T. Koyanbayev ,&nbsp;Ivan M. Kukushkin ,&nbsp;Igor M. Bolshinsky ,&nbsp;Kenneth J. Bateman","doi":"10.1016/j.jnucmat.2025.155801","DOIUrl":"10.1016/j.jnucmat.2025.155801","url":null,"abstract":"<div><div>The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy.</div><div>This paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20% in terms of ²³⁵U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix.</div><div>The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155801"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783936","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":"On the degradation of Young's modulus of irradiated U-10Mo","authors":"Chaoyue Jin ,&nbsp;Zhexiao Xie ,&nbsp;Luning Chen ,&nbsp;Xingdi Chen ,&nbsp;Jing Zhang ,&nbsp;Shurong Ding ,&nbsp;Xiaobin Jian","doi":"10.1016/j.jnucmat.2025.155799","DOIUrl":"10.1016/j.jnucmat.2025.155799","url":null,"abstract":"<div><div>Theoretical analysis of the four-point bending experimental results from the reference has demonstrated that the effective Young's modulus of heavily-irradiated U-10Mo fuel undergoes a significant reduction. However, the underlying mechanisms need to be fully elucidated. In this study, the irradiation-induced thermo-mechanical coupling behaviors of monolithic fuel plates are first numerically investigated by employing the fuel skeleton creep-based volumetric growth strain model and the porosity-related macroscale creep rate model for the contained U-10Mo fuel foils. The predicted average thicknesses for the bending specimens from several fuel plates align well with the experimental measurements, validating the adopted models, algorithms and the obtained macroscale porosity values for irradiated U-10Mo fuel. The values of effective Young's modulus of U-10Mo fuel after different levels of irradiation are identified through the subsequent direct simulations of the four-point bending tests, with the numerically acquired macroscale mechanical responses of irradiated U-10Mo specimens matching the experimental data. After eliminating the effects of fuel porosity, it is found that the values of Young's modulus of dense U-10Mo fuel skeleton decrease with increasing fission density or macroscale porosity, thereby becoming a primary contributor to the degradation of the effective Young's modulus of irradiated U-10Mo fuel. Furthermore, mathematical models for the Young's modulus of irradiated U-10Mo fuel skeleton at room temperature are developed as functions of fission density and macroscale porosity, respectively. The predicted results indicate that the von Mises stress will significantly decrease and the equivalent creep strains might have a distinct increase, when the degradation of Young's modulus of fuel skeleton is incorporated. This work provides a foundation for the high-precise modeling of the irradiation-induced thermo-mechanical behaviors of the U-10Mo-based fuel elements or assemblies.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155799"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143814979","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":"Effects of pre-existing vacancy-type dislocation loop on the irradiation resistance in FeNiCoCrCu high-entropy alloy","authors":"Weidong Song ,&nbsp;Zhonghao Huo ,&nbsp;Lijun Xiao ,&nbsp;Lifang Wang ,&nbsp;Jun Chen ,&nbsp;Meizhen Xiang","doi":"10.1016/j.jnucmat.2025.155797","DOIUrl":"10.1016/j.jnucmat.2025.155797","url":null,"abstract":"<div><div>Several high-entropy alloys (HEAs) show considerable promise as structural materials for nuclear energy applications, owing to their exceptional mechanical properties and radiation resistance. However, there is limited understanding of how pre-existing dislocation loops in these HEAs influence their radiation resistance. This study employs molecular dynamics (MD) simulations to investigate the influence of pre-existing dislocation loops on the irradiation resistance of FeNiCoCrCu HEA, focusing on the evolution of point defects, the formation of defect clusters, and the interactions between dislocation loops and point defects during irradiation process. Consequently, the interaction between irradiation-induced point defects and pre-existing dislocation loops leads to an increase in the number of point defects and defect clusters. This is attributed to the reduction in formation energy of point defects by the dislocation loop, which promotes their generation and alters their distribution, thereby inhibiting recombination between them. Point defects migrate toward the dislocation loop under stress field interactions, with the loop exhibiting preferential absorption of interstitial atoms over vacancies, serving as predominant sinks for defect accumulation. Furthermore, the presence of dislocation loops mitigates elemental segregation. During irradiation, dislocation loops absorb vacancies via positive climb and interstitials via negative climb. Meanwhile, the position and shape of the dislocation loop undergo changes, and its length increases. Notably, the FeNiCoCrCu HEA demonstrates enhanced resistance to pre-existing vacancy loop interactions compared to pure Ni, as evidenced by fewer irradiation-induced defects and reduced dislocation loop evolution post-irradiation. These findings elucidate the intricate interplay of defect dynamics in irradiated HEAs and provide critical insights for designing radiation-tolerant HEA systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155797"},"PeriodicalIF":2.8,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783938","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 behavior and mechanical properties of SiC/SiC composite joints with Y2O3-Al2O3-SiO2 interlayer under high-temperature steam environments at 1200 °C","authors":"Shaobo Yang ,&nbsp;Chenxi Liang ,&nbsp;Jiali Li ,&nbsp;Yujie Ma ,&nbsp;Sijie Kou ,&nbsp;Juanli Deng ,&nbsp;Bo Chen ,&nbsp;Shangwu Fan","doi":"10.1016/j.jnucmat.2025.155798","DOIUrl":"10.1016/j.jnucmat.2025.155798","url":null,"abstract":"<div><div>The study investigated the corrosion behavior and mechanical performance of SiC/SiC composite joints with Y₂O₃-Al₂O₃-SiO₂ (YAS) interlayers under high-temperature steam environments at 1200 °C. Under low-flow conditions, partial disruption of Si-O and Al-O bonds in the YAS glass network reduced crosslinking, forming an aluminosilicate protective layer that inhibited further corrosion. Prolonged exposure led to Y³⁺ migration and accumulation, resulting in Y₂Si₂O₇ precipitation and growth. High-flow conditions caused a thinner glass layer, continuous longitudinal cracks, and more severe erosion and dissolution of the YAS glass due to higher steam velocity. Despite these degradations, the joints exhibited satisfactory performance, maintaining shear strengths of about 40 ± 2 MPa after 15 h of low-flow exposure and about 36 ± 5 MPa after 5 h of high-flow exposure. These findings demonstrate that YAS interlayers provide excellent corrosion resistance and mechanical stability as a sealant for nuclear-grade SiC/SiC.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155798"},"PeriodicalIF":2.8,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768652","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}