Journal of Nuclear Materials最新文献

{"title":"The influence of alloying elements on the He ion irradiation damage behavior of Ni0.9M0.1 alloys (M=Cr, Mo, W)","authors":"Ying Li ,&nbsp;Yulin Wei ,&nbsp;Min Liu ,&nbsp;Yajuan Zhong ,&nbsp;Ping Peng","doi":"10.1016/j.jnucmat.2025.156145","DOIUrl":"10.1016/j.jnucmat.2025.156145","url":null,"abstract":"<div><div>He ion irradiation was performed on pure Ni and binary Ni-based alloys (Ni_0.9Cr_0.1, Ni_0.9Mo_0.1 and Ni_0.9W_0.1) at 700 °C to systematically investigate the effects of alloying elements on He bubble and dislocation evolution. Transmission electron microscopy (TEM) revealed a gradual decrease in the average He bubble size in the sequence: pure Ni &gt; Ni_0.9Cr_0.1&gt; Ni_0.9Mo_0.1 &gt; Ni_0.9W_0.1, accompanied by an inverse trend in bubble number density. Distinct differences in dislocation loop evolution were also observed, and Ni_0.9Cr_0.1 exhibited the largest loops among the alloys. All irradiated samples exhibited irradiation-induced hardening, as characterized by nanoindentation. Density functional theory (DFT) calculations were conducted to elucidate the role of alloying elements in He atom diffusion and bubble nucleation. Mo and W suppress bubble coarsening by enhancing lattice atom stability and increasing He migration barriers. In contrast, Cr forms attractive interactions with He via orbital hybridization and electron cloud overlap, acting as additional trapping centers to promote He bubble nucleation. In pure Ni, rapid He migration and diminished lattice stability induce defect enrichment in the local He-affected regions, further forming small dislocation loops that coalesce into dislocation lines or networks. In alloys, reduced He mobility and enhanced interstitial solubility stabilize He defects, driving continuous defect growth. Specifically, Cr-He interactions stabilize He to form large-sized dislocation loops, while Ni_0.9W_0.1 alloys exhibit smaller loops due to higher lattice atom stability. This study provides valuable insights for the design and optimization of irradiation-resistant Ni-based alloys for advanced nuclear systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156145"},"PeriodicalIF":3.2,"publicationDate":"2025-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144988949","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":"Lifetime of the stainless steel inner case of the waste disposal package under gamma irradiation","authors":"Jan Stoulil ,&nbsp;David Dobrev ,&nbsp;Richard Bureš ,&nbsp;Jaroslav Fojt ,&nbsp;Vlastimil Miler","doi":"10.1016/j.jnucmat.2025.156146","DOIUrl":"10.1016/j.jnucmat.2025.156146","url":null,"abstract":"<div><div>In this work, the kinetics of the dissolution of the passive layer of 316 L stainless steel was investigated in the environment of bentonite Černý Vrch (BCV). The composition of the passive layer was characterized using Mott-Schottky curves and XPS. The dissolution kinetics was studied using potentiostatic polarization in the bentonite pore solution and bentonite slurry, as well as exposures in the bentonite pore solution followed by analysis of the released ions using ICP-OES. For the dissolution of Cr from the passive layer, the presence of oxidizing radiolysis products is a key factor, while for the dissolution of Fe, the excitation of the charge carriers over a band gap by radiation is a more significant factor. Exposures in a bentonite pore solution allowed us to determine the kinetics of spontaneous dissolution of the passive layer, which in the repository geometry reaches an equilibrium state within 10³ years. Potentiostatic measurements allowed the steady state dissolution of the passive layer to be rapidly reached, which is at 2.23 × 10^–2 nm.a^-1. Based on the data obtained, numerical simulations were carried out and subsequently extrapolated the lifetime of the inner case of the container, which significantly exceeds the required 10⁶ years.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156146"},"PeriodicalIF":3.2,"publicationDate":"2025-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144997069","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":"Synergistic corrosion mechanisms in ceramic and ceramic-iron joint coatings under lithium-lead exposure and gamma-ray irradiations","authors":"Hnin Lai Lai Wai ,&nbsp;Kento Shirota ,&nbsp;Ryosuke Norizuki ,&nbsp;Khiem Do-Duy ,&nbsp;Hiroyuki Noto ,&nbsp;Teruya Tanaka ,&nbsp;Juro Yagi ,&nbsp;Takuya Nagasaka ,&nbsp;Takumi Chikada","doi":"10.1016/j.jnucmat.2025.156132","DOIUrl":"10.1016/j.jnucmat.2025.156132","url":null,"abstract":"<div><div>In the fusion reactor advanced blanket concepts, considerable attention has always been directed towards the materials employed for the coolants/breeders because the blanket components will be exposed to a harsh environment, including neutron/gamma-ray irradiation and high-temperature tritium breeders. Multifunctional ceramic coatings have been studied to reduce hydrogen isotope permeation, electrical conductivity, and corrosion of structural materials. For the sustainable development of a liquid blanket system, the coatings require not only permeation reduction but also resistance to irradiation damage, which is invisible, accumulative, and can be able to unexpected failure within a short period. In this study, the synergistic corrosion mechanisms of Li-Pb compatibility for the blanket materials were investigated through Li-Pb exposure tests under gamma-ray irradiation. A liquid-phase method and a hot-isostatic pressing were applied for the formation of ceramic-iron joint coatings on the inner wall of a steel tube. In static and flowing Li-Pb exposure tests under gamma-ray irradiation, the joint coatings and multi-layer coatings indicated superior corrosion resistance.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156132"},"PeriodicalIF":3.2,"publicationDate":"2025-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932589","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":"Mechanical properties, 475°C embrittlement and irradiation resistance of FeCrAl-Y2O3 ODS alloy with Ti and V alloying additions","authors":"T. Stasiak ,&nbsp;J.J. Jasiński ,&nbsp;M. Zieliński ,&nbsp;M. Wilczopolska ,&nbsp;A. Kosińska ,&nbsp;W. Chmurzyński ,&nbsp;K. Mulewska ,&nbsp;R. Diduszko ,&nbsp;M. Chmielewski ,&nbsp;M. Kowal ,&nbsp;J. Rzempołuch ,&nbsp;K. Ciporska ,&nbsp;Ł. Kurpaska ,&nbsp;J. Jagielski","doi":"10.1016/j.jnucmat.2025.156131","DOIUrl":"10.1016/j.jnucmat.2025.156131","url":null,"abstract":"<div><div>FeCrAl-based ODS alloys are promising for Generation IV nuclear reactor components since they present remarkable properties, such as high mechanical strength and good irradiation resistance. However, there is still relatively little information on the impact of minor alloying elements, e.g., Ti or V, and manufacturing conditions on the properties of these alloys. In this paper, two FeCrAl-based alloys with the minor addition of Y₂O₃, Ti, and V were studied. The investigation focuses on the mechanical properties at high temperatures (up to 800 °C), resistance to so-called “475 °C embrittlement” (α and α’ phase separation), and irradiation resistance. The conducted tensile tests show promising mechanical properties at room temperature, i.e., the yield strength is 893±23 MPa, and the ultimate tensile strength is 993±10 MPa. The tensile tests at high temperatures also confirm good properties, e.g., at 600 °C the yield strength is 490±8 MPa, the ultimate tensile strength is 526±14 MPa, while the total elongation is 17±4 %. The results reveal that adding vanadium increases the strength of the FeCrAl ODS alloys. The conducted annealing at 475 °C for 1000 h shows limited hardening and almost no change in elongation of the alloy with the addition of vanadium. The performed ion irradiation experiments show good irradiation resistance of both tested alloys, manifesting by limited hardening and stable oxide precipitates. The results confirm the promising properties of studied materials for application in the nuclear industry, particularly in Generation IV reactors. Moreover, the results show the possibility of tuning properties by adding alloying elements, such as vanadium.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156131"},"PeriodicalIF":3.2,"publicationDate":"2025-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932590","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 role of electron correlations and spin-orbit interaction in predicting electrical and heat transport of uranium monocarbide","authors":"Urszula D. Wdowik ,&nbsp;Małgorzata Sternik ,&nbsp;Dominik Legut","doi":"10.1016/j.jnucmat.2025.156093","DOIUrl":"10.1016/j.jnucmat.2025.156093","url":null,"abstract":"<div><div>Electrical and heat transport in UC, a potential fuel material for generation IV nuclear reactors, is investigated within density functional theory incorporating strong local Coulomb and spin-orbit interactions. The localization of 5<em>f</em> electrons is tuned by varying the Coulomb repulsion interaction parameter from 0 to 3 eV. We demonstrate that both strong electron correlation and spin-orbit coupling effects are crucial for realistic modeling of the electron-phonon scattering process, which is a driving mechanism of the electrical and heat transport in UC. Partially localized 5<em>f</em> states, described by a moderate value of the on-site Coulomb repulsion parameter of 1.5 eV, together with spin-orbit interaction reproduce experimental resistivity and thermal conductivity in UC with exceptionally good accuracy in a wide temperature range extending from 300 to 1900 K. The present theoretical approach can potentially be used to eliminate the discrepancy between theory and experiment, as well as to predict the thermoelectric properties of other actinide-based fuel materials for modern nuclear reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156093"},"PeriodicalIF":3.2,"publicationDate":"2025-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144988948","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":"Fabrication of alloy 625/316H cladded component using hot isostatic pressing: Corrosion and mechanical property evaluation in molten chloride salts","authors":"Minsung Hong ,&nbsp;Seong Sik Hwang ,&nbsp;Chaewon Kim ,&nbsp;Seong-Jun Ha ,&nbsp;Joon-Yeop Kwon ,&nbsp;Gyeong-Hoi Koo","doi":"10.1016/j.jnucmat.2025.156130","DOIUrl":"10.1016/j.jnucmat.2025.156130","url":null,"abstract":"<div><div>This study introduced the fabrication of hot isostatic pressing (HIP)-cladded Alloy 625 (A625) / type 316H stainless steel (SS316H) specimens, and evaluated the microstructural characteristics, mechanical properties, and corrosion resistance in comparison with their wrought counterpart. The HIP process effectively suppressed the formation of secondary phases and minimized internal defects, resulting in enhanced mechanical performance, particularly in terms of strength and ductility—key properties for high-temperature structural applications. In molten NaCl–KCl–MgCl₂ salt environments, HIP-treated samples exhibited improved corrosion resistance compared to the wrought materials. However, when both the cladding and substrate were simultaneously exposed to the molten salt, severe galvanic corrosion occurred in the SS316H substrate due to differences in electrochemical potential.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156130"},"PeriodicalIF":3.2,"publicationDate":"2025-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932591","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":"Formation mechanisms of multilayered cobalt deposits on 304 stainless steel in primary cooling systems of pressurized water reactor nuclear power plants","authors":"Jian Deng ,&nbsp;Guolong Wang ,&nbsp;Xiaochen Lv ,&nbsp;Qiuyang Du ,&nbsp;Shiyu Tan ,&nbsp;Zeyong Lei ,&nbsp;Jieheng Lei","doi":"10.1016/j.jnucmat.2025.156124","DOIUrl":"10.1016/j.jnucmat.2025.156124","url":null,"abstract":"<div><div>Radioactive corrosion products, specifically ⁵⁸Co and ⁶⁰Co, accumulate on the surface of components in pressurized water reactors (PWRs), posing significant challenges to nuclear plant safety. In this study, the microstructure and chemical composition of the surface layer of 304 stainless steel (304SS) exposed to cobalt-containing boron/lithium water at a high temperature of 573 K for 5, 10, 15, 20, and 25 days were investigated. The cobalt deposition behaviour and mechanisms were analysed via material characterization techniques, E–pH diagrams, Gibbs free energy calculations, and analysis of the preference energies of metal cations in crystallographic structures. The results revealed that after 15 days of soaking, three distinct cobalt deposition layers formed on the 304SS surface. The outer layer (∼65 nm) consisted of Co₃O₄, the middle layer (∼16 nm) consisted of CoFe₂O₄, and the inner layer (∼4 nm) consisted of CoCr₂O₄. The composition of these layers was relatively independent of the soaking time. CoCr₂O₄ primarily formed through migration of Co²⁺ from solution, coprecipitation with dissolved Cr³⁺ from the specimen surface or ion exchange with FeCr₂O₄ and NiCr₂O₄. CoFe₂O₄ formed through coprecipitation with dissolved Fe³⁺ or ion exchange with NiFe₂O₄ and Fe₃O₄. Co₃O₄ was derived from oxidative decomposition of Co(OH)₂ at high temperatures. This study provides key insights into the formation mechanisms of cobalt deposition layers on 304SS in PWRs and provides a theoretical reference for optimization of primary water chemical environmental parameters, improvement of structural materials, and selection of decontamination methods during operation or decommissioning.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156124"},"PeriodicalIF":3.2,"publicationDate":"2025-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144922046","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":"Raman spectroscopic investigation of UCl4","authors":"Brodie S. Barth ,&nbsp;Teagan F.M. Sweet ,&nbsp;Avery L. Wood ,&nbsp;Peter C. Burns ,&nbsp;Andrew Miskowiec ,&nbsp;Tyler L. Spano","doi":"10.1016/j.jnucmat.2025.156129","DOIUrl":"10.1016/j.jnucmat.2025.156129","url":null,"abstract":"<div><div>Uranium chloride salts are a proposed fuel source for molten salt reactors (MSRs). However, despite their relevance to nuclear energy, they remain understudied, in part because of their air and moisture sensitivity. This work provides the first Raman spectra of UCl₄ from 45 to 3200 cm⁻¹ collected with 532 and 785 nm excitation sources as well as the assignment of 10 identified peaks to their respective vibrational modes. These Raman bands are compared to those for isostructural ThCl₄, the compositionally related UCl₃, and the computed Raman bands for UCl₄ from local vibrational mode analysis. The observed spectrum of UCl₄ is in accord with that of ThCl₄ and the computed spectra of UCl₄. We posit that the observed differences between the spectra of UCl₄ and UCl₃ are useful in differentiating these species for applications such as in situ monitoring of MSR operations.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156129"},"PeriodicalIF":3.2,"publicationDate":"2025-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144997068","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":"Synergistic trapping and desorption effects of vacancy and heterogeneous interface on helium in W-TiC alloys by first-principles calculations","authors":"ShuLong Wen ,&nbsp;XiangYang Wang ,&nbsp;Min Pan ,&nbsp;HuiQiu Deng","doi":"10.1016/j.jnucmat.2025.156128","DOIUrl":"10.1016/j.jnucmat.2025.156128","url":null,"abstract":"<div><div>The synergistic trapping effect of vacancies and the W/TiC interface on helium, as well as the desorption behaviors of helium at the W/TiC interface with and without vacancies were investigated using first-principles calculations. The stability of interface structures with multiple interfacial orientations was assessed through the work of separation. Building on this, the thermodynamic stability of vacancies near the interface and their trapping effect on helium were examined. The presence of these vacancies significantly hindered the absorption of helium at the interface. Helium initially accumulated at vacancies near the interface, forming small Vac-<em>n</em>He clusters. As the number of captured helium atoms increased, helium became more prone to capture by the interface, thereby inhibiting the growth of Vac-<em>n</em>He clusters. The nearest-neighbor and second-nearest-neighbor vacancies near the interface together with the interface formed community and co-solvent structures, respectively. These structures can effectively promote the interface's ability to trap helium within bulk W, compared to the pure interface. The desorption temperatures of helium at the W/TiC interface with and without vacancy range from 1060 K to 1580 K and from 900 K to 1100 K, respectively. The presence of vacancies significantly elevated the desorption temperature of helium at the interface. These research findings contribute to a deeper understanding of the nucleation and growth mechanisms of helium bubbles, as well as the desorption of helium at the W/TiC interface.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156128"},"PeriodicalIF":3.2,"publicationDate":"2025-08-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144921975","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":"Lowered tritium diffusivity by decreased site density along Li2TiO3 (001) twin boundary","authors":"Zhonghua Lu ,&nbsp;Yanli Shi ,&nbsp;Yuchen Liu ,&nbsp;Huanhuan Liu ,&nbsp;Xiuling Wang ,&nbsp;Cong Zhang ,&nbsp;Gaoyuan Wang ,&nbsp;Jianqi Qi ,&nbsp;Tiecheng Lu","doi":"10.1016/j.jnucmat.2025.156127","DOIUrl":"10.1016/j.jnucmat.2025.156127","url":null,"abstract":"<div><div>Li₂TiO₃ is a prominent tritium breeder material with significant potential in the design of nuclear fusion reactors. Given that tritium recovery is closely related to its behavior at grain boundaries (GBs), understanding the dynamics is critical for optimizing the tritium release performance of the breeder material. In this work, we investigate the diffusion of tritium at the Li₂TiO₃ (001) twin GB with density functional theory and kinetic Monte Carlo. Tritium segregation sites are identified and local migration paths between these sites are analyzed. The segregation energies are in the range of -0.43 ∼ 0.06 eV. The activation energies for tritium migration along the <strong><em>a</em></strong>-axis and the <strong><em>b</em></strong>-axis of the GB are 0.54 eV and 0.56 eV, respectively. The diffusion coefficient for tritium along the GB is estimated to be 1.97 × 10^–7 exp(-0.45 eV/K_B<em>T</em>) m²/s. The GB affects tritium diffusivity mainly by raising the activation energy along the <strong><em>b</em></strong>-axis, which consequently lowers the diffusivity by less than one order of magnitude within the 300∼1200 K temperature range compared to the bulk. The decreased tritium site density at the GB compared to bulk is proposed to be the reason for the modified diffusion behaviors by the GB.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"617 ","pages":"Article 156127"},"PeriodicalIF":3.2,"publicationDate":"2025-08-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144913625","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}