Journal of Nuclear Materials最新文献

{"title":"Hydride growth behaviors in lamellar U-2Nb alloy","authors":"Wenliang Xu ,&nbsp;Dawu Xiao ,&nbsp;Wenyuan Wang ,&nbsp;Denglei Chen ,&nbsp;Sheng Zhang ,&nbsp;Min Wu ,&nbsp;Zili Yuan ,&nbsp;Rongguang Zeng ,&nbsp;Hefei Ji ,&nbsp;Fan Liu ,&nbsp;Tao Fa ,&nbsp;Bin Su ,&nbsp;Xinchun Lai","doi":"10.1016/j.jnucmat.2025.155793","DOIUrl":"10.1016/j.jnucmat.2025.155793","url":null,"abstract":"<div><div>Hydride growth behaviors are known to be governed by microstructure and stress, but the interplay of the two factors remain unclear. In this work, the growth behaviors of uranium hydride (UH₃) in lamellar U-2Nb alloy were systematically investigated. The growth of UH₃ in U-2Nb samples was controlled by cathodically hydrogen charging with different current densities. The hydrides were categorized into three types based on their growth rate and nucleation sites: (1) blisters at α-U lamellae, (2) fast growth families around inclusions, and (3) fishbone-like families at prior α-U grain boundaries (GBs). Surface and cross-sectional morphologies of these hydrides were examined by focused-ion-beam (FIB) milling and scanning electron microscope (SEM). The results showed that, governed by lamellar microstructure, the hydride propagation along the lamellar direction (LD) was observed throughout the hydriding progress. Meanwhile, the preference of spherical hydrides was enhanced by increasing strain energy. The volume expansion induced tensile fields and cracks were found dominating the formation of acicular UH₃, and the hydrides could penetrate far into the matrix through the α-U lamellae. Furthermore, the hydride growth behaviors and their corresponding hydriding mechanisms in lamellar U-2Nb alloy, covering microscopic and early macro scale, are elucidated in this work.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":""},"PeriodicalIF":2.8,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143776526","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 potential emerging issue concerning repair welding of out-of-core PWR components involving tritium exposure and 3He retention","authors":"F.A. Garner ,&nbsp;M.N. Gussev ,&nbsp;M. Song ,&nbsp;G.S. Was","doi":"10.1016/j.jnucmat.2025.155774","DOIUrl":"10.1016/j.jnucmat.2025.155774","url":null,"abstract":"<div><div>This paper identifies a potential but previously unrecognized risk of helium-induced embrittlement and cracking during repair welding of out-of-core pressurized water reactor (PWR) components exposed to tritium-contaminated coolant. While previous weldability concerns centered on ⁴He accumulation in neutron-irradiated alloys located within the in-core or near-core regions, new measurements show that ³He generated by tritium decay can accumulate in out-of-core components. Because hydrogen isotopes readily diffuse to grain boundaries and become trapped there, significant ³He generation at grain boundaries may lead to cracking during weld repairs. Initial data from far-below-core PWR flux thimble tubes confirm the presence of ³He levels above known cracking thresholds for repair welds. These findings indicate that out-of-core regions should be considered when defining safe weld repair windows in reactors operating for 60–100 years.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":""},"PeriodicalIF":2.8,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143776525","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 Impact of Cascade Overlap on the Long-Term Evolution of Helium Bubbles in Ferritic/Martensitic Steels","authors":"Xin-Hua Yan,&nbsp;Lu Sun,&nbsp;Teng Xie,&nbsp;Zhen-Feng Tong","doi":"10.1016/j.jnucmat.2025.155809","DOIUrl":"10.1016/j.jnucmat.2025.155809","url":null,"abstract":"<div><div>A comprehensive investigation into the evolution mechanisms of helium bubbles under irradiation is critical for understanding the swelling and embrittlement effects induced by helium bubbles in reduced activation ferritic/ martensitic (RAFM) steels. New findings have been made regarding the role of cascade overlap in the long-term evolution of helium bubbles in RAFM steels under irradiation, by combining Cluster Dynamics (CD) and Molecular Dynamics (MD) statistical trends. The Transmission Electron Microscopy (TEM) observations have been compared with the CD simulations, with the simulations including cascade overlap showing better consistency with the TEM observations. A comparative analysis of the simulation results from the CD model with helium bubble cascade overlap and the CD model without cascade overlap revealed that the cascade overlap significantly changes the size-density distribution of helium bubbles, the helium-to-vacancy (He/V) ratio, and helium bubble pressure. The atomically small helium bubbles initiating nucleation undergo complete dissolution following cascade overlap, which is a critical factor in reducing helium bubbles density. The current simulation results suggest that helium bubbles undergoing cascade overlap are more likely to stably nucleate through a stable configuration where the He/V ratio remains unchanged after multiple cascade overlaps and helium bubbles with a high He/V ratio in their vicinity, which alter the He/V ratio and pressure state of the helium bubbles. Cascade overlap leads to an elevation in vacancy and interstitial helium concentrations, and the formation of high He/V ratio helium bubbles, which together contribute to the expansion of helium bubble size.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155809"},"PeriodicalIF":2.8,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143807730","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":"Unpacking model inadequacy: The quantification of silver release from TRISO fuel by considering empirical and mechanistic approaches","authors":"Somayajulu L.N. Dhulipala ,&nbsp;Pierre-Clément A. Simon ,&nbsp;Paul A. Demkowicz ,&nbsp;Jacob A. Hirschhorn ,&nbsp;Stephen R. Novascone","doi":"10.1016/j.jnucmat.2025.155795","DOIUrl":"10.1016/j.jnucmat.2025.155795","url":null,"abstract":"<div><div>Increasing adoption of the proposed tristructural isotropic (TRISO) particle fuel for both advanced and existing reactors makes it critical to assess and address any uncertainties and inadequacies of TRISO fission product release models. Model inadequacy stems from simplifications made to the computational model when compared to the experiments. The modeling and simulation efforts conducted using the BISON fuel performance code, along with the experimental campaigns carried out under the Advanced Gas Reactor Fuel Development and Qualification Program, afford a unique opportunity to conduct a rigorous modeling inadequacy assessment within the Bayesian uncertainty quantification (UQ) framework. This study compares the standard Bayesian framework against the Kennedy-O'Hagan (KOH) framework, which explicitly represents modeling inadequacy, in regard to UQ for TRISO silver release models. For this purpose, both the traditional Arrhenius equation fitted to experimental data and the more advanced lower-length-scale (LLS)-informed model, which considers microstructure information, are independently considered. Applying the inverse UQ process on the AGR-2 and -3/4 datasets revealed modeling inadequacy to be the most dominant source of uncertainty. Experimental noise uncertainty is also significant; however, model parameter uncertainty can be considered negligible. Interestingly, both the Arrhenius equation and the LLS-informed model demonstrated similar levels of modeling inadequacy. For the forward predictive UQ, the KOH framework improved both the accuracy and quality of quantified uncertainties in comparison to the standard Bayesian framework. This is true for both the Arrhenius equation and the LLS-informed model. In comparing these modeling approaches, both demonstrated similar performance at the engineering scale, while the LLS-informed model expectedly outperformed the Arrhenius equation at the mesoscale. These conclusions highlight the importance of explicitly accounting for modeling inadequacy in the UQ process, and reinforce the need for continuous refinement of physics-based models in order to address the modeling inadequacy.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155795"},"PeriodicalIF":2.8,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143790847","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 Effect of Proton Irradiation on In-situ Corrosion Behavior of Zr-Sn-Nb Zirconium Alloy under Simulated PWR Water Condition","authors":"Tianxu Chen,&nbsp;Jingjing Liao,&nbsp;Shaoyu Qiu","doi":"10.1016/j.jnucmat.2025.155802","DOIUrl":"10.1016/j.jnucmat.2025.155802","url":null,"abstract":"<div><div>In this study, we conducted a 30-day in-situ electrochemical corrosion investigation on proton-irradiated N36 zirconium alloy samples. Both potentiodynamic polarization and electrochemical impedance spectroscopy (EIS) methods were utilized to examine changes in electrochemical properties. Additionally, a long-term corrosion study was performed. Through the incorporation of microscopic characterizations, including X-ray diffraction (XRD) and transmission electron microscopy (TEM), we explored the influence of irradiation on the corrosion behavior and microstructure of the N36 alloy. Our findings demonstrate that proton irradiation reduced the corrosion rate of the N36 alloy and led to the formation of a denser oxide layer. Notably, as the irradiation dose increased, the beneficial effects became more pronounced. This improvement is likely attributed to the increased stability of the tetragonal zirconia (t-ZrO₂) phase, facilitated by defect generation, Sn/Nb doping in the zirconia, and significant irradiation-induced hardening. These factors promote the formation of a dense phase during the initial stages of corrosion.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155802"},"PeriodicalIF":2.8,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143790845","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":"Speciation mapping of the oxidation layer on aged uranium dioxide using scanning transmission x-ray spectromicroscopy","authors":"Nic Cicchetti ,&nbsp;Alexander Ditter ,&nbsp;Joseph I. Pacold ,&nbsp;Zurong Dai ,&nbsp;Scott B. Donald ,&nbsp;Brandon W. Chung ,&nbsp;M. Lee Davisson ,&nbsp;Artem V. Gelis ,&nbsp;David K. Shuh","doi":"10.1016/j.jnucmat.2025.155792","DOIUrl":"10.1016/j.jnucmat.2025.155792","url":null,"abstract":"<div><div>In this study, UO₂ was aged in humid air and prepared as a thin section using a focused ion beam (FIB) instrument. The specimen was measured using synchrotron radiation spectromicroscopy techniques at the Beamline 11.0.2 STXM end station of the Advanced Light Source (ALS). Non-negative matrix factorization (NMF) methods were used to identify and map three component x-ray absorption near-edge structure (XANES) spectra in the oxygen K-edge data, revealing a surface layer of U₄O₉ with a thickness of 206 ± 21 nm, and the bulk of the sample remaining as UO₂. Uranium N_4,5-edge XANES spectromicroscopy supports these results. The diffusion-controlled parabolic rate constant for UO₂ oxidation to U₄O₉ was calculated from the observed layer thickness and compared to literature values. Complementary transmission electron microscopy (TEM) was used to image the sample and identify the phases present in various regions, confirming the STXM results.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155792"},"PeriodicalIF":2.8,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143843404","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":"First-principles prediction of thermodynamic and mechanical properties of ZrCr2 under extreme conditions","authors":"Hongling Zhou ,&nbsp;Baifeng Luan ,&nbsp;An Yan ,&nbsp;Xiaoling Yang ,&nbsp;Congqing Liu ,&nbsp;Xuyang Liu ,&nbsp;Chunrong Xu ,&nbsp;Chao Sun ,&nbsp;Haibo Ruan ,&nbsp;Weijiu Huang ,&nbsp;Korukonda L. Murty","doi":"10.1016/j.jnucmat.2025.155807","DOIUrl":"10.1016/j.jnucmat.2025.155807","url":null,"abstract":"<div><div>The cubic C15 and hexagonal C14/C36 phases of ZrCr₂ are typical phases formed in the newly developed Cr-coated Zr alloy cladding, and their formation and transformation significantly affect the system's service performance. However, their crystal structures and associated thermodynamic and mechanical properties under high temperatures and pressures are not fully elucidated. This work comprehensively explores the phase stability, thermodynamic, and mechanical properties of these three ZrCr₂ polymorphs over a temperature range from 0 to 2500 K and pressure from 0 to 30 GPa using first-principles calculations. The calculated lattice parameters agree well with experimental values. It is confirmed that C15 is the stable phase at low temperatures, C36 serves as an intermediate phase, and C14 is relatively prevalent at high temperatures. Above 10 GPa, C14 is no longer stable in the temperature range studied. Temperature and pressure significantly influence the thermodynamic and mechanical properties of the three ZrCr₂ phases. All three phases are ductile, but their ductility decreases as temperature rises and pressure decreases, accompanied by volume expansion, suggesting that ZrCr₂ formation during a loss of coolant accident could significantly increase the risk of cracking. These results advance our understanding of the thermodynamic and mechanical behavior of the ZrCr₂ phases under high temperatures and pressures, providing essential insights for designing high-performance Cr-coated Zr alloy cladding in nuclear engineering.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155807"},"PeriodicalIF":2.8,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799584","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":"Ion irradiation effects on the mechanical behavior of vanadium micropillars","authors":"Lei Li ,&nbsp;Zhixi Zhu ,&nbsp;Jiaohui Yan ,&nbsp;Shang Chen ,&nbsp;Qiuhong Zhang ,&nbsp;Yaguang Dong ,&nbsp;Xun Guo ,&nbsp;Ke Jin ,&nbsp;Yunfei Xue","doi":"10.1016/j.jnucmat.2025.155784","DOIUrl":"10.1016/j.jnucmat.2025.155784","url":null,"abstract":"<div><div>The present work reports the impact of proton irradiation on the mechanical behavior of vanadium micropillars with various diameters at room temperature. Irradiation induced strengthening appears more significant with increasing pillar size, and the critical diameter to reach size-independent strength is strongly reduced after irradiation to 0.2 dpa. Although all unirradiated pillars deform uniformly, the deformations of the irradiated pillars thicker than 1 μm are concentrated within the top 30 % region, where most irradiation defects are wiped by deformation dislocations, leading to softening and the consequent unique deformation localization behavior.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155784"},"PeriodicalIF":2.8,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143850583","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":"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}