Journal of Nuclear Materials最新文献_第2页

Micro-scale signature for suppressing fragile fracture in high entropy alloys under irradiation 辐照下抑制高熵合金脆性断裂的微尺度特征

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-17 DOI: 10.1016/j.jnucmat.2025.155834

Peng-wei Wang , Babafemi Malomo , Shu-quan Chang , Liang Yang

{"title":"Micro-scale signature for suppressing fragile fracture in high entropy alloys under irradiation","authors":"Peng-wei Wang , Babafemi Malomo , Shu-quan Chang , Liang Yang","doi":"10.1016/j.jnucmat.2025.155834","DOIUrl":"10.1016/j.jnucmat.2025.155834","url":null,"abstract":"<div><div>Compared with traditional alloys, high entropy alloys (HEAs) have better resistance to irradiation embrittlement and hardening, which continue to gain significant attention as promising high-end structural materials, but up until now, the underpinnings of suppressing brittle failure are yet to be revealed, limiting their application. Hence, this study proposes a molecular dynamics framework that can apprehend the evolutions of nano-scale dislocations and micron-sized shear bands from a microstructural evolution-energetics standpoint to elucidate deformation mechanisms in FeNiCrCuAl HEAs under irradiation. Accordingly, prototypic models (0 dpa, 0.02 dpa and 0.2 dpa) of the HEA, indicated an ultimate tensile strength at equivalent strain point of 4.7 % but as strengths declined with the progression of strain, multiple irradiations provoked intense atomic-dislocation interactions by which higher dislocation densities stimulated high-energy dislocation intersects for an amplified work-hardening effect. The evolutions of dislocation density with variations in average atomic energies precipitated distinctive shear band mechanisms characterized by multiple shear bands propagations along 45° and 135° directions in all of the models, and by the atomic level internal stresses constraint on atomic mobility under negative pressure, lowered atomic energies induced by intensified irradiations evolved a phenomenal cross-blocking effect of fewer multiple propagating shear bands to indicate higher ultimate tensile strength and enhanced resistance to fragile failure in the HEAs. Thus, by capturing the dislocation-shear band mechanism under irradiated energy potential landscape, the correlation between micro-scale structural evolutions and mechanical behavior was established in unraveling the fragile failure phenomenon in HEAs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155834"},"PeriodicalIF":2.8,"publicationDate":"2025-04-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143854815","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

Development of boron-based thermal neutron absorbers for short-pulsed MW-class neutron sources – how can the pre-decoupling function be enhanced? 用于短脉冲毫瓦级中子源的硼基热中子吸收剂的研制——如何增强预解耦功能？

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-16 DOI: 10.1016/j.jnucmat.2025.155779

Toshifumi Okutomi , Makoto Teshigawara , Sota Sato , Stephen Gallimore , Robert Bewley , Motoki Ooi , Masahide Harada , Shigeru Kuramoto

引用次数: 0

Statistical study of displacement cascades in Ni and FeNiCr alloys: Understanding the influence of potential and composition on primary damage modeling Ni和FeNiCr合金中位移级联的统计研究：了解电位和成分对初级损伤建模的影响

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-14 DOI: 10.1016/j.jnucmat.2025.155832

Adithya Nair , Charlotte S. Becquart , Christophe Domain , Andrée De Backer

{"title":"Statistical study of displacement cascades in Ni and FeNiCr alloys: Understanding the influence of potential and composition on primary damage modeling","authors":"Adithya Nair , Charlotte S. Becquart , Christophe Domain , Andrée De Backer","doi":"10.1016/j.jnucmat.2025.155832","DOIUrl":"10.1016/j.jnucmat.2025.155832","url":null,"abstract":"<div><div>This work investigates the interaction between high-energy particles and metals, focusing on the primary irradiation damage through extensive molecular dynamics simulations. The cascades are simulated using empirical interatomic potentials and cover an extensive range of energies (above and below the sub cascade threshold), ranging from 0.5 keV to 120 keV. These potentials are characterized using properties associated with point defects, surface energy, stacking fault energy, threshold displacement energy, and Quasi-Static Drag (QSD). The data obtained from the simulations are analyzed using specific descriptors, which helps improve our understanding of the primary damage.</div><div>A database containing approximately 15,000 displacements cascades in both nickel (Ni) and the FeNiCr alloy has been generated by molecular dynamics. To assess these extensive datasets, a variety of statistical methodologies, including MANOVA, ANOVA, k means and correlation matrices, were employed. Utilizing these analytical tools and statistical descriptors, the study investigated the influences of potentials and compositions on defect production in both nickel (with 3 different Ni potentials) and 5 different FeNiCr compositions. A comparative analysis between the outcomes of potential and alloy analyses was conducted to determine the predominant effect.</div><div>Potentials exhibit varied effects, particularly post-fragmentation energy, influencing cascade fragmentation and mono defects. Alloy compositions showcase differing defect production patterns, with Ni generating more defects, while alloys produce an elevated number of mono-interstitials and interstitial clusters. Notably, the study highlights the impact of the Ni fragmentation energy, identifying differing effects below and above this threshold, with a pronounced influence on interstitials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155832"},"PeriodicalIF":2.8,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143873902","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

Corrosion behavior of Ti-Grade2 dissolver material in nitric acid containing fluoride ions ti - 2级溶解剂材料在含氟硝酸中的腐蚀行为

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-13 DOI: 10.1016/j.jnucmat.2025.155833

R. Priya , K. Kaliraj , S. Ningshen

{"title":"Corrosion behavior of Ti-Grade2 dissolver material in nitric acid containing fluoride ions","authors":"R. Priya , K. Kaliraj , S. Ningshen","doi":"10.1016/j.jnucmat.2025.155833","DOIUrl":"10.1016/j.jnucmat.2025.155833","url":null,"abstract":"<div><div>The Ti-grade2 is used as the candidate dissolver material for reprocessing the nuclear-spent fuels in the aqueous-based Plutonium Uranium Recovery by EXtraction (PUREX) reprocessing method. Fluoride is used to enhance the dissolution of high-burn-up fuel. However, fluoride is known to severely degrade the corrosion behaviour of commercially pure Ti (CP-Ti) in reprocessing nitric acid. The present work focuses on understanding the corrosion behaviour of Ti-grade2 dissolver material in nitric acid medium containing fluoride ions with and without complexing agent Al(NO<sub>3</sub>)<sub>3</sub> at 11.5 M and 1 M HNO<sub>3</sub> by electrochemical and boiling nitric acid studies. The potentiodynamic polarization results revealed inferior corrosion resistance with active corrosion potential and higher passive current density in 11.5 M HNO<sub>3</sub> + 0.05 M NaF compared to 11.5 M HNO<sub>3</sub>. Moreover, the deterioration of corrosion resistance was more pronounced with increasing temperature and fluoride concentration in 11.5 M HNO<sub>3</sub>. Spontaneous passivation behavior was observed under all conditions tested in this acid concentration. In contrast, the presence of fluoride in 1 M HNO<sub>3</sub> induced an active-passive transition, characterized by a negative shift in corrosion potential and an increase in passive current density. Corrosion mitigation of titanium was found to be effective in a nitric acid medium containing fluoride and a complexing agent. A significantly higher corrosion rate was observed in 1 M HNO<sub>3</sub> + 0.05 M NaF compared to 11.5 M HNO<sub>3</sub> + 0.05 M NaF during the boiling nitric acid test. SEM, XPS, and AFM analyses supported these findings. The corrosion mechanism by which fluoride influences the corrosion resistance of CP-Ti in both 1 M and 11.5 M nitric acid concentrations was proposed. Additionally, this study provides valuable insights for the use of CP-Ti as a dissolver material in a nuclear reprocessing environment involving nitric acid and fluoride.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155833"},"PeriodicalIF":2.8,"publicationDate":"2025-04-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143837835","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

Assessing the viability of a new subsize tensile specimen geometry for evaluation of structural nuclear and additively manufactured materials 评估用于评估核结构材料和快速成型材料的新型亚尺寸拉伸试样几何形状的可行性

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-12 DOI: 10.1016/j.jnucmat.2025.155831

David Collins , Maxim Gussev , Stephen Taller , T.S. Byun , Caleb Massey

{"title":"Assessing the viability of a new subsize tensile specimen geometry for evaluation of structural nuclear and additively manufactured materials","authors":"David Collins , Maxim Gussev , Stephen Taller , T.S. Byun , Caleb Massey","doi":"10.1016/j.jnucmat.2025.155831","DOIUrl":"10.1016/j.jnucmat.2025.155831","url":null,"abstract":"<div><div>The use of subsize specimens in nuclear materials testing has been a subject of ongoing interest due to radiation safety concerns and resource conservation needs. It is also of interest in additive manufacturing (AM), also known as 3D-printing, as subsize specimens can more accurately represent the behavior of the small, intricate geometries often produced using AM. A novel, extremely small geometry, called the Subsize Teeny (SST) was recently developed and is of interest for implementation. Given its extraordinarily small size and the complexities associated with subsize specimen testing, adequate vetting of this geometry is necessary to ensure data quality. A variety of unirradiated nuclear structural materials were tested in the SST geometry and compared against the well-established SSJ3 geometry. In addition, two case studies implementing the SST as a screening geometry for AM materials were also conducted. The question of SST viability was found to be highly nuanced and will often be dependent on the context or application in question. It was determined, however, that the SST is a largely invalid geometry for exceptionally coarse-grained materials or in cases where the physical defect volume equals or exceeds 0.1 % or where the specimen machining parameters result in significant surface alterations. On the other hand, it was determined that the SST may be employed with confidence if the test material is nearly or totally free of physical defects, isotropic, demonstrates homogeneous plastic deformation, and possesses a fine-grained, nearly or totally homogeneous microstructure with at least twelve slip systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155831"},"PeriodicalIF":2.8,"publicationDate":"2025-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143843402","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

Polarizable force fields for the structural and thermophysical properties of molten actinide chlorides 熔态锕系氯化物的结构和热物理性质的极化力场

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-11 DOI: 10.1016/j.jnucmat.2025.155822

Giovanni Pireddu , Agustin Salcedo , Hugo Sauzet , Sylvie Delpech , David Lambertin , Timothée Kooyman

{"title":"Polarizable force fields for the structural and thermophysical properties of molten actinide chlorides","authors":"Giovanni Pireddu , Agustin Salcedo , Hugo Sauzet , Sylvie Delpech , David Lambertin , Timothée Kooyman","doi":"10.1016/j.jnucmat.2025.155822","DOIUrl":"10.1016/j.jnucmat.2025.155822","url":null,"abstract":"<div><div>New nuclear technologies could involve the extensive use of molten salts, including actinide halides. Despite their importance, several practical challenges limit experimental measurements, resulting in knowledge gaps for structural and thermophysical properties. In this work, new polarizable force fields based on ab initio calculations for the simulation of molten actinide chlorides are introduced. The new force fields are used to compute structural properties, density, heat capacity, and isothermal compressibility of pure actinide molten salts (ThCl<sub>4</sub>, PaCl<sub>3</sub>, NpCl<sub>3</sub>, AmCl<sub>3</sub>, CmCl<sub>3</sub>) at various temperatures. UCl<sub>3</sub> and PuCl<sub>3</sub>, which were parameterized in previous works, are also included. The results are discussed in the context of already existing theoretical and experimental datasets, showing good agreement with the literature. Predictions are extended to systems not considered in previous works. Notably, the results highlight the peculiarity of ThCl<sub>4</sub> compared to actinide trichlorides in terms of structural and thermophysical properties. The new force fields can be used in future works for the simulation of molten salts mixtures containing actinides.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155822"},"PeriodicalIF":2.8,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143854814","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

The role of pre-adsorbed hydrogen in facilitating water release at the Li4TiO4 surface: A first-principles study 预吸附氢在促进 Li4TiO4 表面水释放中的作用：第一原理研究

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-11 DOI: 10.1016/j.jnucmat.2025.155823

Zhonghua Lu , Yanli Shi , Yuchen Liu , Huanhuan Liu , Xiuling Wang , Cong Zhang , Gaoyuan Wang , Jianqi Qi , Tiecheng Lu

{"title":"The role of pre-adsorbed hydrogen in facilitating water release at the Li4TiO4 surface: A first-principles study","authors":"Zhonghua Lu , Yanli Shi , Yuchen Liu , Huanhuan Liu , Xiuling Wang , Cong Zhang , Gaoyuan Wang , Jianqi Qi , Tiecheng Lu","doi":"10.1016/j.jnucmat.2025.155823","DOIUrl":"10.1016/j.jnucmat.2025.155823","url":null,"abstract":"<div><div>Li<sub>4</sub>TiO<sub>4</sub> is recognized as an attractive material for tritium breeding in fusion reactors, owing to its high lithium content. Understanding the process of surface water formation is important for optimizing its tritium release performance. In this work, we conducted first-principles calculations to investigate the transformation of desorbed H<sub>2</sub> to water molecules on the Li<sub>4</sub>TiO<sub>4</sub> surface. This process encompasses the dissociative adsorption of H<sub>2</sub>, the diffusion of hydrogen atoms, the abstraction of a surface oxygen atom to form water molecules, and their subsequent release. The possible adsorption sites, local diffusion pathways, and their corresponding activation energies are identified. We determined that the global energy barrier for the transformation from desorbed H<sub>2</sub> to desorbed H<sub>2</sub>O in the pre-adsorbed hydrogen system is 1.62 eV, approximately 1 eV less than the 2.67 eV observed in the non-adsorbed hydrogen system. This reduction in the energy barrier suggests that pre-adsorbed hydrogen facilitates water release. The decrease in the energy barrier is attributed to the easier formation of oxygen vacancies on the surface in the presence of pre-adsorbed hydrogen. Our results are consistent with experimental observations that pre-adsorbed hydrogen promotes the release of tritiated water.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155823"},"PeriodicalIF":2.8,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143837760","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

Interpretation of dissolution behavior at the surface of uranium-zirconium oxide solid solutions 氧化铀锆固溶体表面溶解行为的解释

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-10 DOI: 10.1016/j.jnucmat.2025.155820

Ryutaro Tonna , Takayuki Sasaki , Yoshihiro Okamoto , Taishi Kobayashi

{"title":"Interpretation of dissolution behavior at the surface of uranium-zirconium oxide solid solutions","authors":"Ryutaro Tonna , Takayuki Sasaki , Yoshihiro Okamoto , Taishi Kobayashi","doi":"10.1016/j.jnucmat.2025.155820","DOIUrl":"10.1016/j.jnucmat.2025.155820","url":null,"abstract":"<div><div>The dissolution behavior of (U,Zr)O<sub>2</sub>, the primary uranium solid phase in the fuel debris from the Fukushima Daiichi nuclear power plant accidents, was investigated thermodynamically and kinetically under atmospheric conditions. Cubic (U,Zr)O<sub>2</sub> samples with a uniform solid solution of Zr were prepared using wet chemistry methods, and static batch immersion tests were conducted. In strongly acidic conditions, where the solubility of U and Zr exceeded their concentrations, congruent dissolution of both elements was observed with (U,Zr)O<sub>2</sub> dissolving at the same rate as UO<sub>2</sub>. In moderately acidic conditions, where the U solubility was higher than its concentration with Zr reaching a steady state at lower solubility, the U dissolution rate from (U,Zr)O<sub>2</sub> decreased compared to UO<sub>2</sub>. In the presence of oxalic acid, with increased Zr solubility due to the formation of complexes, the U dissolution rate from (U,Zr)O<sub>2</sub> did not decrease. This indicates that Zr in (U,Zr)O<sub>2</sub> formed a secondary solid phase on the solid surface under conditions of lower Zr solubility, which in turn suppressed the oxidative dissolution of U.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155820"},"PeriodicalIF":2.8,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143843403","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

On the microstructural evolution and hydriding behavior of dilute Zr-2.5Nb-Y alloys 稀Zr-2.5Nb-Y合金组织演变及氢化行为研究

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-10 DOI: 10.1016/j.jnucmat.2025.155819

Y. Pushpalatha Devi , K.V. Mani Krishna , N. Keskar , J.B. Singh , R.N. Singh

{"title":"On the microstructural evolution and hydriding behavior of dilute Zr-2.5Nb-Y alloys","authors":"Y. Pushpalatha Devi , K.V. Mani Krishna , N. Keskar , J.B. Singh , R.N. Singh","doi":"10.1016/j.jnucmat.2025.155819","DOIUrl":"10.1016/j.jnucmat.2025.155819","url":null,"abstract":"<div><div>Zr-2.5Nb alloy is a critical pressure tube material in pressurized heavy water reactors (PHWRs), as its performance directly influences the operational life and safety of the reactor. However, the formation of hydrides in this alloy can detrimentally affect its mechanical properties. In the present study, yttrium was employed as a dilute addition to mitigate hydride embrittlement. Alloys were prepared with varying yttrium contents of 0.5 to 2 wt %. Yttrium addition resulted in the formation of fine yttria precipitates. The hot deformed microstructures of alloys with Y exhibited significant differences in morphology of the phases and prior β phase fraction when compared to reference Zr2.5Nb alloy. Dilatometry studies indicated that yttrium addition led to a reduction in the β-transus temperature of the alloy. The hydride behavior of the alloys was also examined, showing that yttrium significantly reduced hydride size to <20 μm, compared to a range of 10–150 μm in the absence of yttrium (Zr-2.5Nb). This comprehensive study of the microstructure and hydriding behavior, with the addition of yttrium to the Zr-2.5Nb alloy, suggests that yttrium may be considered for improving the alloy's performance in nuclear applications, in view of the mitigation of hydride embrittlement.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"611 ","pages":"Article 155819"},"PeriodicalIF":2.8,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143821050","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 thermal stability and irradiation resistance of ODS-W/CuCrZr joints by interlayer employ and interface improvement 通过层间和界面改进，提高了ODS-W/CuCrZr接头的热稳定性和耐辐照性

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-04-10 DOI: 10.1016/j.jnucmat.2025.155818

Dang Xu , Changcheng Sang , Kaichao Fu , Ruizhi Chen , Pengqi Chen , Yingwei Lu , Dahuan Zhu , Qiu Xu , Jigui Cheng

{"title":"Enhanced thermal stability and irradiation resistance of ODS-W/CuCrZr joints by interlayer employ and interface improvement","authors":"Dang Xu , Changcheng Sang , Kaichao Fu , Ruizhi Chen , Pengqi Chen , Yingwei Lu , Dahuan Zhu , Qiu Xu , Jigui Cheng","doi":"10.1016/j.jnucmat.2025.155818","DOIUrl":"10.1016/j.jnucmat.2025.155818","url":null,"abstract":"<div><div>To enhance the performance of W/Cu divertor materials under high-temperature and irradiation conditions, this study utilizes oxide dispersion-strengthened tungsten (ODS-W) and CuCrZr alloy as base materials. A tri-layer ODS-W/W-50Cu/CuCrZr joint was fabricated using spark plasma sintering (SPS), incorporating a nanoporous surface treatment on the ODS-W surface and a W-50Cu interlayer between ODS-W and CuCrZr. The effects of the surface treatment and W-50Cu interlayer on the microstructure, mechanical properties, and irradiation resistance of the joints were systematically investigated. Results demonstrate that the nanoporous structure significantly enhances interfacial bonding, achieving a tensile strength of 227.6 MPa and a ductility of 5.82 %. Fracture analysis reveals a transition in failure mode. Fractures shift from the ODS-W/Cu interface to the W-50Cu interlayer, accompanied by a transition from brittle to ductile fracture behavior. The W-50Cu interlayer effectively mitigates the mismatch in thermal expansion and minimizes stress concentrations, thereby enhancing interfacial stability at elevated temperatures while maintaining excellent thermal conductivity and mechanical properties. Under irradiation, the W-50Cu interlayer acts as a “trap”, capturing and neutralizing irradiation-induced defects. This mechanism reduces interfacial damage, mitigates hardening, and improves irradiation stability. These findings establish a framework for optimizing W/Cu divertor material design for high-temperature and irradiation-intensive applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"611 ","pages":"Article 155818"},"PeriodicalIF":2.8,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143833758","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