Journal of Nuclear Materials最新文献_第3页

Corrosion behavior and mechanism of 17–4PH martensitic stainless steel in saturated oxygen static lead-bismuth eutectic at 450 °C 17-4PH马氏体不锈钢在450℃饱和氧静态铅铋共晶中的腐蚀行为及机理

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-07 DOI: 10.1016/j.jnucmat.2025.156155

Guobao Ma , Gen Li , Zhengxin Tang , Xitao Wang , Jinshan He , Hansheng Bao , Xikou He

{"title":"Corrosion behavior and mechanism of 17–4PH martensitic stainless steel in saturated oxygen static lead-bismuth eutectic at 450 °C","authors":"Guobao Ma , Gen Li , Zhengxin Tang , Xitao Wang , Jinshan He , Hansheng Bao , Xikou He","doi":"10.1016/j.jnucmat.2025.156155","DOIUrl":"10.1016/j.jnucmat.2025.156155","url":null,"abstract":"<div><div>17–4PH martensitic stainless steel is used in pressurized water reactors (PWRs) nuclear systems and also shows potential for lead-cooled fast reactors (LFRs), while its resistance to lead-bismuth corrosion is critical. Therefpore, the corrosion behavior of 17–4PH steel in saturated oxygen static lead-bismuth eutectic (LBE) at 450 °C was investigated. It was found that during the initial 3000 h, an incomplete oxide film developed due to the cyclic formation and spalling of oxide layers. By 4500 h, a stable double-layered oxide film was formed, consisting of an outer Fe<sub>3</sub>O<sub>4</sub> layer and an inner FeCr<sub>2</sub>O<sub>4</sub> layer. During internal oxidation, Ni enrichment occurred beneath the internal oxidation layer (IOL)/matrix interface, transforming the body-centered cubic (BCC) martensitic matrix into a face-centered cubic (FCC) Ni-rich phase. The stress induced by the volume change during this phase transition constitutes the predominant driver for premature oxide film spalling. The Ni-rich phases exhibit resistance to oxidation and form isolated islands and flocculent structures within the IOL. Cu accumulates as Cu-rich phases within the IOL and tends to rapidly grow into irregular shapes by adhering to the Ni-rich phase. The flocculent distribution of Cu and Ni within the IOL produced nanoporous regions. These pores acted as diffusion pathways for metallic elements, accelerating oxide film growth after 4500 h. The corrosion mechanism of 17–4PH steel in liquid LBE is discussed based on these observations.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156155"},"PeriodicalIF":3.2,"publicationDate":"2025-09-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045195","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}

引用次数: 0

Influence of damage rate on irradiation hardening behavior in proton-irradiated Fe-Cu alloy 损伤速率对质子辐照Fe-Cu合金辐照硬化行为的影响

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-05 DOI: 10.1016/j.jnucmat.2025.156148

Yusuke Noshi , Moe Ishiwaki , Ryoya Ishigami , Kazuhiro Yasuda , Ken-ichi Fukumoto

{"title":"Influence of damage rate on irradiation hardening behavior in proton-irradiated Fe-Cu alloy","authors":"Yusuke Noshi , Moe Ishiwaki , Ryoya Ishigami , Kazuhiro Yasuda , Ken-ichi Fukumoto","doi":"10.1016/j.jnucmat.2025.156148","DOIUrl":"10.1016/j.jnucmat.2025.156148","url":null,"abstract":"<div><div>To investigate the effects of damage rate on irradiation-induced hardening and microstructural evolution, proton irradiation experiments were conducted on Fe-0.05 wt.%Cu binary alloy samples over a wide range of ion fluxes spanning three orders of magnitude. The samples were then subjected to nano-indentation tests and microstructural observations. Nano-indentation hardness tests revealed greater hardening at lower damage rates, even under the same damage level. TEM observation showed that larger but fewer dislocation loops formed under low damage rate conditions, while high damage rates produced smaller but denser dislocation loops. STEM/EDS analysis showed that as the damage rate decreased, the number density and diameter of Cu-rich precipitates increased, and the Cu concentration therein increased. Irradiation hardening contribution was estimated by invoking the dispersed-barrier hardening model and the Russell-Brown model indicating that the effect of damage rate on irradiation hardening controlled by CRPs is dominant. Since kinetic modeling simulation showed enhanced vacancy mobility at lower damage rates and enhanced Cu-rich precipitate formation, it supported the experimental results. Our findings suggest that damage rate has a significant effect on irradiation hardening behavior, and that low damage rate significantly contributes to the nucleation and growth process of Cu-rich precipitates by thermal diffusion during prolonged irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156148"},"PeriodicalIF":3.2,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045194","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}

引用次数: 0

Atomistic simulation of xenon bubble re-solution at dislocations versus in bulk UO2 due to thermal spike 由于热尖峰，位错处氙气泡再溶解的原子模拟与块状UO2的比较

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-05 DOI: 10.1016/j.jnucmat.2025.156154

L. Yang , B.D. Wirth

{"title":"Atomistic simulation of xenon bubble re-solution at dislocations versus in bulk UO2 due to thermal spike","authors":"L. Yang , B.D. Wirth","doi":"10.1016/j.jnucmat.2025.156154","DOIUrl":"10.1016/j.jnucmat.2025.156154","url":null,"abstract":"<div><div>The re-solution rate of xenon (Xe) bubbles in irradiated uranium dioxide (UO<sub>2</sub>) is a critical parameter related to fission gas bubble evolution and fission gas release. Molecular dynamics (MD) simulations have been used to understand the effect of spatial location near a dislocation, in addition to gas density and temperature, on the re-solution for nanometric Xe bubbles induced by thermal spikes at a ½<110>{100} edge dislocation or a ½<110> screw dislocation, as well in bulk UO<sub>2</sub>. As well, these MD simulations also investigate the effect of bubble shape on re-solution at the edge dislocation and our results show that the re-solution for a bubble at the dislocation has a weak dependence on the spike track direction and the bubble shape. Interestingly, the average value of re-solution from a Xe bubble located near a dislocation is close to that observed in bulk UO<sub>2</sub>. Xe re-solution in the UO<sub>2</sub> matrix is dependent on gas density and temperature, in addition to bubble size. A pressurized bubble has a stronger resistance to thermal spikes than equilibrium bubbles with a similar size. As well, re-solution evidently increases with increasing temperature from 800 to 1500 K. We propose an improved exponentially saturating function to predict re-solution as a function of gas density, bubble size and temperature based on the MD simulation results obtained for Xe bubble re-solution due to thermal spikes in UO<sub>2</sub>.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156154"},"PeriodicalIF":3.2,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045199","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}

引用次数: 0

Nanoscale clustering and fission product segregation in irradiated annular U-10Zr fuel 辐照环形铀- 10zr燃料中的纳米级聚类和裂变产物偏析

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-05 DOI: 10.1016/j.jnucmat.2025.156150

Arnold Pradhan , Sohail Shah , Daniele Salvato , Jason Harp , Luca Capriotti , Mukesh Bachhav , Indrajit Charit , Tiankai Yao

引用次数: 0

Mechanical behaviors of Cr coating on Zircaloy tubes prepared by pulsed laser deposition 脉冲激光沉积锆合金管表面Cr涂层的力学行为

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-04 DOI: 10.1016/j.jnucmat.2025.156149

Ziqi Wei , Bo Li , Lijuan Cui , John Andrew Kane Jovellana , Zideng Wang , Zongda Yang , Huilong Yang , Sho Kano , Toshiyasu O , Hiroaki Abe

{"title":"Mechanical behaviors of Cr coating on Zircaloy tubes prepared by pulsed laser deposition","authors":"Ziqi Wei , Bo Li , Lijuan Cui , John Andrew Kane Jovellana , Zideng Wang , Zongda Yang , Huilong Yang , Sho Kano , Toshiyasu O , Hiroaki Abe","doi":"10.1016/j.jnucmat.2025.156149","DOIUrl":"10.1016/j.jnucmat.2025.156149","url":null,"abstract":"<div><div>Chromium-coated Zircaloy cladding is one promising candidate for accident-tolerant fuel (ATF), yet its mechanical reliability under loading remains insufficiently understood. This study aims to investigate the cracking behavior and fracture resistance of a thin nanocrystalline/amorphous chromium (Cr) coating fabricated by pulsed laser deposition (PLD) under hoop tensile strain. A 350-nm-thick Cr bilayer coating was deposited onto Zircaloy-4 substrates, and its hoop tensile response was evaluated using the advanced expansion due to compression (A-EDC) test. Key mechanical properties, including Young’s modulus (227 GPa), hardness (13.3 GPa), and residual stress (-1450 MPa), were measured by nanoindentation and X-ray diffraction (XRD). Detailed analyses by scanning electron microscopy (SEM) revealed the evolution of transverse and slanted cracks under increasing strain, with a superior first cracking strain of 3.2 %. A crack density-strain relationship was established and used to estimate the fracture strength (5.8 GPa) and interfacial shear strength (980 MPa) via a shear-lag model. The findings highlight the mechanical advantages of the nanocrystalline/amorphous structure, including grain refinement strengthening and the ductile-buffering effect of the amorphous layer. This study provides a new mechanical perspective for advancing the Cr-coated Zircaloy claddings for ATF applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156149"},"PeriodicalIF":3.2,"publicationDate":"2025-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145045197","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}

引用次数: 0

Leaching of B-Si-glass by water in the presence of bentonite 膨润土存在下b -硅玻璃的水浸出

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-04 DOI: 10.1016/j.jnucmat.2025.156133

Victor Malkovsky, Sergey Yudintsev, Maximilian Niсkolsky

{"title":"Leaching of B-Si-glass by water in the presence of bentonite","authors":"Victor Malkovsky, Sergey Yudintsev, Maximilian Niсkolsky","doi":"10.1016/j.jnucmat.2025.156133","DOIUrl":"10.1016/j.jnucmat.2025.156133","url":null,"abstract":"<div><div>Incorporation of high-level radioactive waste (HLW) into borosilicate glass is used on an industrial scale in most countries developed large-scale nuclear power engineering with closed fuel cycle. The vitrified HLW in a steel canister is planned to be disposed of in an underground repository. The bentonite clay serves as an important protective barrier and can significantly influence the durability of the borosilicate glass when groundwater penetrates the canister due to corrosion of its walls or mechanical damage resulting from rock movement caused by rock bursts or increased seismic activity. An experimental study of the leaching of borosilicate glass with simulators of HLW components (U, REEs, corrosion products) by water in the presence of bentonite has been carried out. The effect of the initial mass ratio of water to bentonite on the leaching rate of the glass is analyzed. Colloidal forms of glass leaching products are revealed by filtering the leach solution through membranes of different pore sizes. The elemental content of the filtrates was determined by inductively coupled plasma mass spectroscopy (ICP MS). The surface and cross section of the glass after leaching were studied using scanning electron microscopy (SEM) combined with energy dispersive X-ray analysis (EDX). The influence of the mass ratio of water and bentonite on the leaching of borosilicate glass can be noticeable at the beginning of the process. The presence of bentonite affects the colloidal form of radionuclide simulants in the leach products, in particular, the uranium-bearing colloid in the leach solution decreases radically. The outer surface of the altered glass layer is significantly enriched with aluminum in the presence of bentonite.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156133"},"PeriodicalIF":3.2,"publicationDate":"2025-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145019608","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}

引用次数: 0

First-principles investigation of cerium and neodymium diffusion in BCC chromium and vanadium via vacancy-mediated transport 铈和钕在BCC铬和钒中通过空位介导的传输扩散的第一性原理研究

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-03 DOI: 10.1016/j.jnucmat.2025.156125

Shehab Shousha , Benjamin Beeler , Larry K. Aagesen , Geoffrey L. Beausoleil II , Maria A. Okuniewski

{"title":"First-principles investigation of cerium and neodymium diffusion in BCC chromium and vanadium via vacancy-mediated transport","authors":"Shehab Shousha , Benjamin Beeler , Larry K. Aagesen , Geoffrey L. Beausoleil II , Maria A. Okuniewski","doi":"10.1016/j.jnucmat.2025.156125","DOIUrl":"10.1016/j.jnucmat.2025.156125","url":null,"abstract":"<div><div>Lanthanide transport plays a crucial role in the performance and longevity of metallic nuclear fuels. This study examines the diffusion behavior of Ce and Nd—two major fission products—in body-centered cubic (BCC) Cr and V, which are potential liner or coating materials for mitigating fuel-cladding chemical interactions (FCCI). Using density functional theory (DFT) calculations and self-consistent mean-field (SCMF) analysis, the vacancy-mediated diffusion coefficients are evaluated. Our findings reveal that Ce and Nd act as oversized solutes and are strongly bound to vacancies in BCC Cr and V, with diffusivities in Cr significantly lower than in V and in hexagonal closed-packed (HCP) Zr, as investigated in our previous work. The activation energies for Ce and Nd diffusion are 3.39 and 3.32 eV, respectively, in BCC Cr, and 2.56 and 2.33 eV, respectively, in BCC V. Analysis of vacancy drag and partial diffusion coefficient ratios indicates a strong tendency for lanthanide enrichment at vacancy sinks in BCC Cr, and to a lesser extent in BCC V, with this effect persisting up to the melting point in Cr and remaining substantial for Nd in V at high temperatures. Under irradiation, the increase in vacancy concentration is expected to enhance lanthanide transport, potentially accelerating interactions at liner-cladding interfaces. Although BCC Cr exhibits relatively low lanthanide diffusivities under equilibrium conditions, the expected segregation tendencies under irradiation suggest that Zr liners may be a more favorable option. Further investigations using rate theory, cluster dynamics, and phase-field modeling are required to quantitatively assess the performance of these materials in reactor environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156125"},"PeriodicalIF":3.2,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145019473","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}

引用次数: 0

Post-irradiation annealing study on low-alloy steels: Determining the contributions of matrix damage and clustering to hardening 低合金钢辐照后退火研究：确定基体损伤和聚类对硬化的贡献

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-02 DOI: 10.1016/j.jnucmat.2025.156144

A.J. Cackett , H. Wilcox , A. Kite , J. Ferriday , N. Riddle , P.D. Styman

{"title":"Post-irradiation annealing study on low-alloy steels: Determining the contributions of matrix damage and clustering to hardening","authors":"A.J. Cackett , H. Wilcox , A. Kite , J. Ferriday , N. Riddle , P.D. Styman","doi":"10.1016/j.jnucmat.2025.156144","DOIUrl":"10.1016/j.jnucmat.2025.156144","url":null,"abstract":"<div><div>In this work, post-irradiation annealing (PIA) at temperatures up to 425 °C was carried out on two low-Cu steels with varying Mn content that were irradiated at high flux to doses up to ∼0.1 dpa. Atom probe tomography (APT) was used to characterise the number, size, and composition of clusters in the as-irradiated material and select PIA samples, and the results compared alongside irradiation-induced change in hardness. For both steels it was found that as annealing temperature increased there was dissolution of Mn and Si from the clusters into the surrounding matrix. The number density and volume fraction of clusters were found to decrease with increasing annealing temperature. As expected from particle hardening models, the change in hardness due to irradiation was proportional to the square root of cluster volume fraction normalised by cluster size for the PIA material. The as-irradiated material, however, deviated from this trend. It is proposed that the observed increase above the expected hardening for the as-irradiated material is attributed with unstable matrix damage, which cannot be directly characterised by APT.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156144"},"PeriodicalIF":3.2,"publicationDate":"2025-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144997067","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}

引用次数: 0

Mechanochemical oxidation and dissolution of uranium oxide phases: Implications for nuclear material processing 铀氧化物相的机械化学氧化和溶解：对核材料加工的影响

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-02 DOI: 10.1016/j.jnucmat.2025.156142

Emma L. Markun, Asher B. Motes, Tori Z. Forbes

{"title":"Mechanochemical oxidation and dissolution of uranium oxide phases: Implications for nuclear material processing","authors":"Emma L. Markun, Asher B. Motes, Tori Z. Forbes","doi":"10.1016/j.jnucmat.2025.156142","DOIUrl":"10.1016/j.jnucmat.2025.156142","url":null,"abstract":"<div><div>Mechanochemical synthetic methods are often highlighted as efficient, low-waste, and low-temperature processes, and they could improve the sustainability of current spent nuclear fuel recycling methods. In the present study, we offer an in-depth analysis of the mechanochemical oxidations of UO<sub>2</sub> and U<sub>3</sub>O<sub>8</sub> with Na<sub>2</sub>O<sub>2</sub>/H<sub>2</sub>O<sub>2</sub> and Na<sub>2</sub>CO<sub>3</sub>/H<sub>2</sub>O<sub>2</sub> for use in spent nuclear fuel recycling schemes. Solid-state and solution characterization of the mechanochemical products indicates the formation of water-soluble U(VI) triperoxide and peroxocarbonate products in >85 % yields. We show that uranium can be recovered from the resulting aqueous solutions of the mechanochemical triperoxide and peroxocarbonate products via aging or acidification to form uranyl oxyhydroxide or uranyl peroxide hydrates. An evaluation of the efficiency of the milling experiments shows that the oxidation reactions are enhanced by mechanical processing but may not be mechanically driven. Addition of select fission analogues suggests that fission product partitioning will be similar to that reported for the CARBEX process.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156142"},"PeriodicalIF":3.2,"publicationDate":"2025-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145019607","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}

引用次数: 0

Enhanced precipitation in IN617 induced by ion irradiation at elevated temperatures 离子辐照在高温下诱导IN617的增强沉淀

IF 3.2 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-09-02 DOI: 10.1016/j.jnucmat.2025.156143

Yinan Zhang , Zhou Zhou , Jin Huang , GuoHua Xu , Yang Liu , Lei Wang , Fanqiang Meng

{"title":"Enhanced precipitation in IN617 induced by ion irradiation at elevated temperatures","authors":"Yinan Zhang , Zhou Zhou , Jin Huang , GuoHua Xu , Yang Liu , Lei Wang , Fanqiang Meng","doi":"10.1016/j.jnucmat.2025.156143","DOIUrl":"10.1016/j.jnucmat.2025.156143","url":null,"abstract":"<div><div>The high-temperature irradiation response of Ni-based superalloys remains a critical knowledge gap for advanced nuclear reactor applications. In this study, the microstructural evolution and mechanical property changes in IN617 alloy under Ni<sup>2</sup><sup>+</sup> ion irradiation up to 10 dpa at 600 °C and 800 °C were investigated <em>via</em> combined electron microscopic characterizations and nanoindentation. Microstructural characterizations reveal that irradiation reduces the size and volume fraction of carbides, while promoting γ' phase precipitation. The enhanced γ' precipitation can be attributed to the pronounced atomic diffusion induced by irradiation. Furthermore, the dislocation density in the sample irradiated at 800 °C is much lower than that at 600 °C due to thermal annihilation. Both γ' precipitation and increased dislocation density contribute to the higher nanoindentation hardness in the irradiated samples. These findings establish fundamental relationships between irradiation temperature, precipitate stability, and mechanical degradation in IN617 alloy.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156143"},"PeriodicalIF":3.2,"publicationDate":"2025-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144997066","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}

引用次数: 0