Journal of Nuclear Materials最新文献

{"title":"Atomistic simulation of the interactions of radiation damage and grain boundaries in austenitic FeCrNi alloy","authors":"Dongpeng Hua ,&nbsp;Qiaosheng Xia ,&nbsp;Jincheng Li ,&nbsp;Qing Zhou ,&nbsp;Yeran Shi ,&nbsp;Yuxuan Zhu ,&nbsp;Bida Zhu ,&nbsp;Wenting Ye ,&nbsp;Xiaofei Yu ,&nbsp;Haifeng Wang","doi":"10.1016/j.jnucmat.2025.156234","DOIUrl":"10.1016/j.jnucmat.2025.156234","url":null,"abstract":"<div><div>Austenitic FeCrNi (Fe-20Cr-10Ni) alloy is considered one of the potential structural materials for fourth generation nuclear reactors due to its excellent mechanical properties, corrosion resistance, and radiation resistance. In its study of radiation damage, grain boundary (GB) design is one of the effective strategies to improve its radiation resistance, as GBs can significantly reduce radiation damage by absorbing radiation-induced defects. However, the understanding of the relationship between GB characteristics and radiation resistance is still unclear. This study investigated the cascade collision process of six FeCrNi bi-crystals with different GBs through atomistic simulation, aiming to investigate the defect absorption characteristics of different types of GBs in FeCrNi alloy. The results indicate that compared to the single crystal, the presence of GBs in the bi-crystals can effectively help absorb radiation-induced point defects, thereby suppressing the formation of a large number of clusters and the evolution of dislocation loops and stacking fault tetrahedra. The defect absorption efficiency is determined by the excess energy of GBs, and a universal logarithmic relationship exists between these two parameters. After radiation, there is a clear \"U-shaped\" relationship between the reduced yield strain/stress and the excess energy of GB. In addition, the rearrangement of displaced atoms in GB induced by radiation leads to GB migration and the formation of new GB facets on the {111} plane. Besides, the interaction between radiation-induced point defects and GBs leads to Ni segregation and Cr depletion in the GBs. This study not only contributes to a deeper understanding of the interaction between radiation-induced defects and GBs, but also provides guidance for breaking through the radiation resistance limit of FeCrNi alloys through anti-radiation GB engineering.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156234"},"PeriodicalIF":3.2,"publicationDate":"2025-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321934","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-based equilibrium Monte Carlo simulations of Cr-rich cluster stability in Fe–Cr–Al alloys: Effects of Al content and vacancies on the transition from promotion to suppression","authors":"Y. Abe ,&nbsp;A. Kubo ,&nbsp;S. Ukai ,&nbsp;T. Tsuru","doi":"10.1016/j.jnucmat.2025.156221","DOIUrl":"10.1016/j.jnucmat.2025.156221","url":null,"abstract":"<div><div>Understanding the formation and stability of Cr-rich clusters (CrRCs) is essential for developing accident-tolerant oxide dispersion-strengthened Fe–Cr–Al alloys. This study employs first-principles-based equilibrium Monte Carlo simulations to quantitatively evaluate the effects of Al concentration and irradiation-induced vacancies on CrRC formation and atomic-scale structural behavior. Statistical metrics such as cluster size distribution, radial distribution functions, and short-range order parameters were used. With increasing Al content, CrRC formation exhibited a non-monotonic trend, peaking near 12 at.% Al. This aligns with prior static energy calculations but additionally captures Al segregation at CrRC–matrix interfaces and, at high Al concentrations, apparent Al incorporation within the cluster region causing structural diffusion and increased lattice displacement, which negatively impacts stability. Vacancies showed little direct effect on CrRC formation but strongly interacted with Al, co-segregating near interfaces. This Al–vacancy co-segregation may influence interface stability and diffusion under irradiation. These findings deepen atomic-scale understanding of CrRC behavior under thermal and irradiation conditions, providing foundational data for designing radiation-resistant Fe–Cr–Al alloys and supporting future mesoscale modeling.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156221"},"PeriodicalIF":3.2,"publicationDate":"2025-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321513","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":"Pre-irradiation effect on deuterium retention after high-flux plasma exposure at two temperatures","authors":"Meng-Qi Zhang ,&nbsp;Hao Yin ,&nbsp;Yi-Wen Sun ,&nbsp;Xiu-Li Zhu ,&nbsp;Long Cheng ,&nbsp;Yue Yuan ,&nbsp;Guang-Hong Lu","doi":"10.1016/j.jnucmat.2025.156229","DOIUrl":"10.1016/j.jnucmat.2025.156229","url":null,"abstract":"<div><div>The effects of 3.5 MeV iron (Fe¹³⁺) ions pre-irradiation on deuterium (D) retention and blistering in recrystallized tungsten were investigated after exposure to low-energy (40 eV) and high-flux (1.8 × 10²² D m⁻² s⁻¹) D plasma at two temperatures (480 K and 630 K) with the high fluence of 2.6 × 10²⁶ m⁻². Surface observations showed a high density of small intragranular blisters on the pristine sample and sparse blisters on pre-irradiated samples at 480 K exposure, while severe surface blisters with maximum sizes reaching several tens of micrometers are present on both pristine and pre-irradiated samples at 630 K exposure. Thermal desorption spectra (TDS) of D indicate different types and amounts of D traps formed in the target sample between the two plasma exposures. Additional high-temperature desorption peaks located above 950 K are present in the TDS of both pristine and pre-irradiated samples at 630 K exposure. Large vacancy clusters are supposed to be trapping sites for the high-temperature desorbed D, and they are most likely the result of the evolution of blistering-induced defects. Total D retention calculation indicates that the pre-irradiation enhances D retention at 480 K exposure, but reduces it at 630 K exposure, suggesting a reversal of the enhancement effect at the higher temperature of 630 K. Since the high-temperature desorption region (900–1100 K) accounts for approximately 80 % of the total D retention—and D desorption in this region is higher in the pristine sample—the reversed enhancement effect is primarily attributed to the source of the high-temperature desorption, namely, large vacancy clusters formed by the evolution of blistering-induced defects. This interpretation is supported by blistering observations, where pre-irradiation suppressed surface blistering and thereby reduced associated defects and large vacancy clusters. This work further reveals the complex influence of pre-irradiation on D behavior in tungsten.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156229"},"PeriodicalIF":3.2,"publicationDate":"2025-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321508","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":"Effect of shot peening on the microstructure, residual stress, and mechanical properties of Inconel 690 alloy","authors":"Shengjie Wang ,&nbsp;Jing Yao ,&nbsp;Shuying Zhen ,&nbsp;Yanli Wang ,&nbsp;Ying Yu ,&nbsp;Chunlei Zhao ,&nbsp;Shilei Li ,&nbsp;Yandong Wang","doi":"10.1016/j.jnucmat.2025.156224","DOIUrl":"10.1016/j.jnucmat.2025.156224","url":null,"abstract":"<div><div>Inconel 690 alloy is a key structural material for heat transfer tubes in pressurized water reactor nuclear power plants, where it endures harsh service conditions. Shot peening (SP) is a widely adopted surface treatment technique to enhance the durability and extend the service life of such components. This study systematically examines the effects of SP process parameters, including shot material, peening intensity, and peening coverage, on the microstructure, residual stress, and mechanical properties of Inconel 690 alloy. The results reveal that SP produces a pronounced gradient microstructure comprising a nanocrystalline surface layer, a plastically deformed sub-surface region, and a deep compressive residual stress field extending to about 240 μm. Moreover, SP induces a gradient compressive residual stress field in Inconel 690 alloy, with the magnitude and penetration depth strongly influenced by shot material, peening intensity, and coverage. The surface modifications significantly improve the mechanical performance: the yield strength increases by up to 62.4 % (from 311 MPa to 505 MPa for the SP-300 % sample), and surface hardness rises by approximately 68.5 % (from 2.990 GPa to 5.038 GPa), albeit with a slight reduction in ductility. Additionally, all SP-treated samples exhibit enhanced wear resistance, as evidenced by reduced specific wear rates due to surface hardening. This work delivers fundamental mechanism understanding of the effects of shot peening on microstructure and mechanical properties, and guides the systematic screening and optimization of SP parameters such as shot material, peening intensity, and coverage.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156224"},"PeriodicalIF":3.2,"publicationDate":"2025-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321524","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Corrosion of EP-450 ferritic-martensitic steel after neutron irradiation and long-term storage in spent fuel pool","authors":"Diana A. Merezhko ,&nbsp;Mikhail S. Merezhko ,&nbsp;Michael P. Short","doi":"10.1016/j.jnucmat.2025.156223","DOIUrl":"10.1016/j.jnucmat.2025.156223","url":null,"abstract":"<div><div>Advanced ferritic-martensitic (F/M) steels are often used in sodium-cooled fast nuclear reactors (SFRs), given their superior resistance to swelling, high-temperature embrittlement, and radiation creep. However, a potential vulnerability of F/M steels lies in their low corrosion resistance to prolonged storage in water-filled spent fuel pools. This poses a significant challenge, especially during fuel storage and reactor decommissioning, increasing the risk of radioactive contamination, safety hazards, and substantial financial costs. Here we show and explain significant corrosion susceptibility of EP-450 ferritic-martensitic steel (1Cr13Mo2NbBV) after irradiation in the BN-350 SFR up to 70 dpa and subsequent long-term (20 years at &lt;50 °C) wet storage in the reactor's spent fuel pool. Corrosion morphologies ranged from oxide films to various forms of localized degradation, including very deep pitting, phase-selective attack, intragranular attack, and interphase corrosion. Deep pits mechanistically form via the preferential dissolution of phase boundaries and the tempered martensitic phase, leading to the eventual detachment of unaffected ferrite grains. The presence and preferential dissolution of the tempered martensitic phase, which is absent in ferritic or austenitic steels, is the dominant factor driving this severe localized corrosion damage in EP-450 steel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156223"},"PeriodicalIF":3.2,"publicationDate":"2025-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321522","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":"Transmission electron microscopy study of the synergistic effects of Mn, Ni and Si on neutron irradiated RPV model steels","authors":"Elvira Oñorbe ,&nbsp;Manuel González ,&nbsp;Mercedes Hernández-Mayoral ,&nbsp;Murthy Kolluri ,&nbsp;Elio D’Agata","doi":"10.1016/j.jnucmat.2025.156215","DOIUrl":"10.1016/j.jnucmat.2025.156215","url":null,"abstract":"<div><div>The microstructure of several reactor pressure vessel (RPV) model steels with varying contents of Ni, Mn and Si was examined using Transmission Electron Microscopy (TEM) before and after neutron irradiation conducted within the Lyra-10 program, up to 0.17 dpa at 286 °C. This study provides new insights into the synergistic effects of these alloying elements on irradiation-induced microstructural, particularly the density, size and distribution of dislocation loops. The results reveal a marked decrease in the density of visible dislocation loops with increasing Ni and Mn content, while the average loop size remains unaffected across all compositions. The effect of Mn seems to be more acute in the decrease of dislocation loop density and the trend is not influenced by variations in Si content. The observed reduction in dislocation loop density is hypothesized to be related to the formation of Mn-Ni-Si-rich solute clusters, which likely compete with loops for point defects or nucleate directly on them, hindering their growth. As a result, many dislocation loops fall below the resolution limit of TEM. These findings, supported by complementary techniques such as Atom Probe Tomography and Small-Angle Neutron Scattering, highlight the critical role of solute content in controlling irradiation-induced microstructures in the studied RPV steels.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156215"},"PeriodicalIF":3.2,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321523","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":"Damage behavior of multi-laser powder bed fusion 316L SS irradiated with He ions at high temperature","authors":"Lin Zhao ,&nbsp;Jianfeng Zhang ,&nbsp;Wenzhao Zhang ,&nbsp;Jinlei Yang ,&nbsp;Juan Hou ,&nbsp;Jianjian Li ,&nbsp;Jun Lin","doi":"10.1016/j.jnucmat.2025.156216","DOIUrl":"10.1016/j.jnucmat.2025.156216","url":null,"abstract":"<div><div>ML-PBF (Multi-laser powder bed fusion) is an additive manufacturing solution for large-scale nuclear components, with the irradiation-induced damage behavior of its fabricated materials being critical to their safe service in nuclear reactors. Helium ion irradiation at high temperature (700 °C) was carried out to investigate the differences in high-temperature helium hardening behavior in ML-PBF 316L SS between single laser forming zone and overlap zone. The results showed that both zones exhibited comparable microstructural features, including cellular and columnar substructures. However, the overlap zone displayed a refined average grain size. Additionally, the dislocation density in the overlap zone was smaller due to in-situ annealing during multi-laser processing, along with smaller nano-oxide particles (20.77 nm) and a higher number density (7.8×10²⁰ m⁻³) compared to the single-laser forming zone. The degree of irradiation hardening increased with increasing irradiation dose in both zones. And the overlap zone demonstrated a lower degree of hardening compared to the single-laser forming zone at the same irradiation dose. The average densities of helium bubbles in the single laser forming zone and overlap zone were 3.54×10²³/m³ and 3.16×10²³/m³, while the average sizes were 2.1 and 2.3 nm, respectively. The average densities of Frank's dislocation loops in both zones were 5.13×10²²/m³ and 5.29×10²²/m³, and the average size was 5.7 and 4.9 nm, respectively. Analysis of the contributions of helium bubbles and dislocation loops to irradiation hardening using the DBH model confirmed the disparity in irradiation hardening behavior between the two zones was primarily attributed to the difference in helium bubble characteristics (including number density and size).</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156216"},"PeriodicalIF":3.2,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321514","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A critical review of the history of fabricating monolithic U-Mo fuel plates","authors":"Jason L․ Schulthess ,&nbsp;Walter․ J․ Williams ,&nbsp;Jan-Fong Jue ,&nbsp;R․ Bruce Nielson ,&nbsp;John Merickel ,&nbsp;Yucheng․ Fu","doi":"10.1016/j.jnucmat.2025.156220","DOIUrl":"10.1016/j.jnucmat.2025.156220","url":null,"abstract":"<div><div>The fabrication of monolithic uranium-molybdenum alloy fuels, specifically those developed for high-performance research and test reactors, began in the early 2000s. The primary fuel form consists of uranium alloyed with 10-wt% molybdenum in a thin foil coated with zirconium and encapsulated in aluminum-6061 cladding. Over the years, the process has evolved with different types of casting, heat treatments, rolling schedules, and cladding applications. This review examines the history of these fabrication processes and their impact on microstructure and fuel-swelling performance.</div><div>Even though various fabrication methods were used, we found little correlation between fabrication variation and fuel swelling. This insensitive relationship between fabrication variation and fuel swelling is primarily due to inhomogeneous microstructures that formed during casting and grain refinement that occurred during rolling. We conclude that the fabrication processes we examined produced similar microstructures, indicating that the fuel microstructure is somewhat insensitive to the fabrication parameters evaluated. However, the relatively small amount of historic data, such as those for grain sizes, limited this analysis. More recently fabricated materials, such as those from ongoing irradiation experiment, Mini-Plate-1 and Mini-Plate-2, were also excluded from this analysis and are intended to be reviewed separately. The findings, that fuel microstructure is somewhat insensitive to the fabrication parameters, do not imply that any fabrication method is acceptable, given the uncertainties in data and fuel-swelling observations. For example, only arc melting and vacuum induction melting casting processes were previously explored in the historic fabrication efforts. The findings should not be extrapolated to other casting processes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156220"},"PeriodicalIF":3.2,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321511","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":"Leaching behaviour of Cs, Sr, and Eu nitrates from Fe-rich polymers: Influence of precursor composition, presence of CeO2 nanoparticles, and the effect of high-dose gamma irradiation","authors":"E.D. Mooren ,&nbsp;A. Cambriani ,&nbsp;S. Van Winckel ,&nbsp;R. Alvarez- Sarandes ,&nbsp;W. Bonani ,&nbsp;L. Fongaro ,&nbsp;G. Beersaerts ,&nbsp;V. Tyrpekl ,&nbsp;T. Cernousek ,&nbsp;S. Schreurs ,&nbsp;W. Schroeyers","doi":"10.1016/j.jnucmat.2025.156214","DOIUrl":"10.1016/j.jnucmat.2025.156214","url":null,"abstract":"<div><div>The long half-lives and potential impact of environmental contamination of radionuclides in liquid nuclear waste make radioactive waste management a major concern for society and the nuclear industry. One prospective solution for efficiently immobilising liquid nuclear waste is the use of Alkali-Activated Materials (AAMs). The immobilisation ability of AAMs largely depends on their composition and the effect of the introduced radionuclides on the structure itself. The purpose of this work is to investigate the leaching behaviour of radioactive simulants such as strontium (Sr), caesium (Cs), and europium (Eu) nitrates from iron-rich slag-derived AAM samples, explore the impact of CeO₂ nanoparticles` presence, as well as the effects of gamma irradiation on the structure and leachability of the contaminants. CeO₂ nanoparticles are investigated for preconcentrating radionuclides from liquid nuclear waste and have been previously used in isolating and adsorbing Eu³⁺. Samples received a cumulative dose of 6.51 MGy of radiation from a cobalt-60 (⁶⁰Co) source. Potential changes in mechanical characteristics and microstructure were investigated by SEM analysis and nanoindentation. The leaching tests revealed significant variability in the release of elements such as Fe, Na, Al, Cs, Sr, Eu, and Ce across different samples, influenced by the slag composition and the presence of CeO₂ nanoparticles. No appreciable variations were observed in the simulated radionuclides or structural element releases from the irradiated and non-irradiated samples. Irradiation did not affect the materials' hardness and had a limited effect on elastic modulus (E). From this perspective, the produced AAMs are suitable candidates for nuclear waste immobilisation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156214"},"PeriodicalIF":3.2,"publicationDate":"2025-10-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145263452","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":"Evidence of radiation induced segregation clustering in binary ferritic model alloys","authors":"M. Vrellou ,&nbsp;A. Nomoto ,&nbsp;P. Pareige ,&nbsp;B. Radiguet","doi":"10.1016/j.jnucmat.2025.156211","DOIUrl":"10.1016/j.jnucmat.2025.156211","url":null,"abstract":"<div><div>This study investigates the microstructural changes and hardening mechanisms in irradiated binary Fe-Mn and Fe-Ni model alloys using atom probe tomography (APT). Mn and Ni are key alloying elements in reactor pressure vessel (RPV) steels, enhancing hardenability and mechanical properties while increasing irradiation sensitivity. Neutron irradiation introduces point defects that lead to solute clustering, impacting the mechanical properties of the materials. APT analysis reveals and quantifies the formation of Mn and Ni clusters, which act as barriers to dislocation motion and contribute to irradiation hardening. This research isolates the individual behavior of Mn and Ni by studying undersaturated binary model alloys, providing detailed insights into the mechanisms of radiation-induced segregation and nanoscale solute clustering and these effect on hardening.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"618 ","pages":"Article 156211"},"PeriodicalIF":3.2,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321515","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}