Journal of Nuclear Materials最新文献

{"title":"Hydride growth behaviors in lamellar U-2Nb alloy","authors":"Wenliang Xu ,&nbsp;Dawu Xiao ,&nbsp;Wenyuan Wang ,&nbsp;Denglei Chen ,&nbsp;Sheng Zhang ,&nbsp;Min Wu ,&nbsp;Zili Yuan ,&nbsp;Rongguang Zeng ,&nbsp;Hefei Ji ,&nbsp;Fan Liu ,&nbsp;Tao Fa ,&nbsp;Bin Su ,&nbsp;Xinchun Lai","doi":"10.1016/j.jnucmat.2025.155793","DOIUrl":"10.1016/j.jnucmat.2025.155793","url":null,"abstract":"<div><div>Hydride growth behaviors are known to be governed by microstructure and stress, but the interplay of the two factors remain unclear. In this work, the growth behaviors of uranium hydride (UH₃) in lamellar U-2Nb alloy were systematically investigated. The growth of UH₃ in U-2Nb samples was controlled by cathodically hydrogen charging with different current densities. The hydrides were categorized into three types based on their growth rate and nucleation sites: (1) blisters at α-U lamellae, (2) fast growth families around inclusions, and (3) fishbone-like families at prior α-U grain boundaries (GBs). Surface and cross-sectional morphologies of these hydrides were examined by focused-ion-beam (FIB) milling and scanning electron microscope (SEM). The results showed that, governed by lamellar microstructure, the hydride propagation along the lamellar direction (LD) was observed throughout the hydriding progress. Meanwhile, the preference of spherical hydrides was enhanced by increasing strain energy. The volume expansion induced tensile fields and cracks were found dominating the formation of acicular UH₃, and the hydrides could penetrate far into the matrix through the α-U lamellae. Furthermore, the hydride growth behaviors and their corresponding hydriding mechanisms in lamellar U-2Nb alloy, covering microscopic and early macro scale, are elucidated in this work.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":""},"PeriodicalIF":2.8,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143776526","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A potential emerging issue concerning repair welding of out-of-core PWR components involving tritium exposure and 3He retention","authors":"F.A. Garner ,&nbsp;M.N. Gussev ,&nbsp;M. Song ,&nbsp;G.S. Was","doi":"10.1016/j.jnucmat.2025.155774","DOIUrl":"10.1016/j.jnucmat.2025.155774","url":null,"abstract":"<div><div>This paper identifies a potential but previously unrecognized risk of helium-induced embrittlement and cracking during repair welding of out-of-core pressurized water reactor (PWR) components exposed to tritium-contaminated coolant. While previous weldability concerns centered on ⁴He accumulation in neutron-irradiated alloys located within the in-core or near-core regions, new measurements show that ³He generated by tritium decay can accumulate in out-of-core components. Because hydrogen isotopes readily diffuse to grain boundaries and become trapped there, significant ³He generation at grain boundaries may lead to cracking during weld repairs. Initial data from far-below-core PWR flux thimble tubes confirm the presence of ³He levels above known cracking thresholds for repair welds. These findings indicate that out-of-core regions should be considered when defining safe weld repair windows in reactors operating for 60–100 years.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":""},"PeriodicalIF":2.8,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143776525","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Low-temperature sintering of (U,Pu)O2 MOX in mild oxidative conditions","authors":"Jacobus Boshoven ,&nbsp;Jean-François Vigier ,&nbsp;Philipp Pöml ,&nbsp;Abibatou Ndiaye ,&nbsp;Bertrand Morel ,&nbsp;Rudy J.M. Konings ,&nbsp;Karin Popa ,&nbsp;Marco Cologna","doi":"10.1016/j.jnucmat.2025.155800","DOIUrl":"10.1016/j.jnucmat.2025.155800","url":null,"abstract":"<div><div>We compare typical reductive sintering conditions for U, Pu mixed oxides (4 h at 1700°C in Ar/6% H₂ + 1200 ppm H₂O) with lower temperature and mildly oxidative conditions (2 h at 1200°C in CO/CO₂ = 1/9) and report on the resulting microstructures and homogeneity. We show that lower temperature and mildly oxidative conditions, without cover gas change, can give close-to stoichiometric, crack-free MOX pellets with a relative density of ∼ 95%, and we propose ways to improve the homogenisations of PuO₂ and UO₂ and increase the grain size.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155800"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783935","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development of the Technological Process for the IGR Reactor's Highly-Enriched Irradiated Uranium-Graphite Fuel Immobilization","authors":"Kuanysh K. Samarkhanov ,&nbsp;Yuliya Yu. Baklanova ,&nbsp;Olga S. Bukina ,&nbsp;Viktor V. Baklanov ,&nbsp;Yerbolat T. Koyanbayev ,&nbsp;Ivan M. Kukushkin ,&nbsp;Igor M. Bolshinsky ,&nbsp;Kenneth J. Bateman","doi":"10.1016/j.jnucmat.2025.155801","DOIUrl":"10.1016/j.jnucmat.2025.155801","url":null,"abstract":"<div><div>The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy.</div><div>This paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20% in terms of ²³⁵U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix.</div><div>The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155801"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783936","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":"Effects of pre-existing vacancy-type dislocation loop on the irradiation resistance in FeNiCoCrCu high-entropy alloy","authors":"Weidong Song ,&nbsp;Zhonghao Huo ,&nbsp;Lijun Xiao ,&nbsp;Lifang Wang ,&nbsp;Jun Chen ,&nbsp;Meizhen Xiang","doi":"10.1016/j.jnucmat.2025.155797","DOIUrl":"10.1016/j.jnucmat.2025.155797","url":null,"abstract":"<div><div>Several high-entropy alloys (HEAs) show considerable promise as structural materials for nuclear energy applications, owing to their exceptional mechanical properties and radiation resistance. However, there is limited understanding of how pre-existing dislocation loops in these HEAs influence their radiation resistance. This study employs molecular dynamics (MD) simulations to investigate the influence of pre-existing dislocation loops on the irradiation resistance of FeNiCoCrCu HEA, focusing on the evolution of point defects, the formation of defect clusters, and the interactions between dislocation loops and point defects during irradiation process. Consequently, the interaction between irradiation-induced point defects and pre-existing dislocation loops leads to an increase in the number of point defects and defect clusters. This is attributed to the reduction in formation energy of point defects by the dislocation loop, which promotes their generation and alters their distribution, thereby inhibiting recombination between them. Point defects migrate toward the dislocation loop under stress field interactions, with the loop exhibiting preferential absorption of interstitial atoms over vacancies, serving as predominant sinks for defect accumulation. Furthermore, the presence of dislocation loops mitigates elemental segregation. During irradiation, dislocation loops absorb vacancies via positive climb and interstitials via negative climb. Meanwhile, the position and shape of the dislocation loop undergo changes, and its length increases. Notably, the FeNiCoCrCu HEA demonstrates enhanced resistance to pre-existing vacancy loop interactions compared to pure Ni, as evidenced by fewer irradiation-induced defects and reduced dislocation loop evolution post-irradiation. These findings elucidate the intricate interplay of defect dynamics in irradiated HEAs and provide critical insights for designing radiation-tolerant HEA systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155797"},"PeriodicalIF":2.8,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783938","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 behavior and mechanical properties of SiC/SiC composite joints with Y2O3-Al2O3-SiO2 interlayer under high-temperature steam environments at 1200 °C","authors":"Shaobo Yang ,&nbsp;Chenxi Liang ,&nbsp;Jiali Li ,&nbsp;Yujie Ma ,&nbsp;Sijie Kou ,&nbsp;Juanli Deng ,&nbsp;Bo Chen ,&nbsp;Shangwu Fan","doi":"10.1016/j.jnucmat.2025.155798","DOIUrl":"10.1016/j.jnucmat.2025.155798","url":null,"abstract":"<div><div>The study investigated the corrosion behavior and mechanical performance of SiC/SiC composite joints with Y₂O₃-Al₂O₃-SiO₂ (YAS) interlayers under high-temperature steam environments at 1200 °C. Under low-flow conditions, partial disruption of Si-O and Al-O bonds in the YAS glass network reduced crosslinking, forming an aluminosilicate protective layer that inhibited further corrosion. Prolonged exposure led to Y³⁺ migration and accumulation, resulting in Y₂Si₂O₇ precipitation and growth. High-flow conditions caused a thinner glass layer, continuous longitudinal cracks, and more severe erosion and dissolution of the YAS glass due to higher steam velocity. Despite these degradations, the joints exhibited satisfactory performance, maintaining shear strengths of about 40 ± 2 MPa after 15 h of low-flow exposure and about 36 ± 5 MPa after 5 h of high-flow exposure. These findings demonstrate that YAS interlayers provide excellent corrosion resistance and mechanical stability as a sealant for nuclear-grade SiC/SiC.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155798"},"PeriodicalIF":2.8,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768652","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":"Impact of helium ion implantation on deuterium plasma induced microstructure evolution and deuterium retention in tungsten","authors":"Honghui Zhang ,&nbsp;Tongjun Xia ,&nbsp;Yongzhi Shi ,&nbsp;Zhengyu Jiang ,&nbsp;Xingyu Ren ,&nbsp;Lisha Liang ,&nbsp;Kaigui Zhu","doi":"10.1016/j.jnucmat.2025.155794","DOIUrl":"10.1016/j.jnucmat.2025.155794","url":null,"abstract":"<div><div>Surface blistering and internal microstructure evolutions as well as deuterium retention in tungsten with helium ion implanted followed by deuterium plasma exposure were investigated. The helium ion implantation was taken with 40 keV with a flux of 1.6 × 10¹⁷ He⁺/(m²s) to a fluence of 6.0 × 10²⁰ He⁺/m² at room temperature. The following deuterium plasma exposure was taken with a flux of 5.96 × 10¹⁹ D/(m²s) at a bias of 100 eV at 340 K. The deuterium plasma exposure was designed with two different durations. One is about 19 h (h) which corresponds a fluence of 4.07 × 10²⁴ D/m², while another is nearly 96 h corresponds a fluence of 2.06 × 10²⁵ D/m². The helium ion implantation itself did not induce surface blister nor detectable internal helium bubble. After subsequent deuterium exposure of 19 h, dense surface blisters appeared on the reference tungsten, while no blister was formed on the helium implanted tungsten, indicating the helium ion implantation can efficiently suppress the surface blistering. However, when the deuterium irradiation time was increased up to 96 h, sparse deuterium blisters appeared on the surface of the helium ion pre-implanted W, indicating D could pass through the helium implantation layer as the exposure time was long enough. TEM results revealed that no bubble can be observed in the reference tungsten only exposed to deuterium plasma, while bubbles can be confirmed in the helium ion pre-implanted tungsten after deuterium irradiation, suggesting that the growth of helium bubbles can be enhanced by the subsequent deuterium plasma exposure. For the deuterium plasma exposure with 19 h, the total deuterium retention in the helium ion pre-implanted tungsten was three times that of the reference tungsten, indicating the helium ion implantation could increase the deuterium retention in tungsten.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155794"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777488","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":"Impact of calcium and pH on ISG alteration at basic pH: Mechanism of formation and transport properties of the gel layer","authors":"Benjamin Cagnon ,&nbsp;Stéphane Gin ,&nbsp;Martiane Cabié ,&nbsp;Damien Daval","doi":"10.1016/j.jnucmat.2025.155796","DOIUrl":"10.1016/j.jnucmat.2025.155796","url":null,"abstract":"<div><div>The dissolution of International Simple Glass (ISG) was investigated at 90 °C, elevated concentration of dissolved silica and in the presence of calcium, with a specific emphasis on basic pH conditions. The leaching solution was labelled with ²⁹Si, ¹⁸O and ⁴⁴Ca in part of the experiments to elucidate the dissolution mechanisms. Based on the isotopic signatures of the gel layer analyzed using Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS), it was concluded that oxygen atoms mostly originate from the solution for all investigated conditions, while silicon atoms almost exclusively originate from the glass. A negative correlation was found between the initial concentration of calcium in solution and the gel layer thickness, suggesting either the formation of a passivating (Si, Ca)-rich layer, a catalytic effect of Ca on the gel densification or a combination of both. In addition, the pH-dependence of the diffusion coefficient of B within the gel was found to be stronger in the basic pH range than in the acidic pH range, which was suggested to originate from the change in coordination of B species at pH₉₀ °_C ∼ 8.5. Overall, these results suggest that in a (Ca, Si)-rich solution at basic pH, the durability of ISG is stronger than previously thought, as the diffusion coefficient of B under such conditions are lower than expected based on literature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155796"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768659","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Effect of Mn replacing Ni on the microstructure and tensile properties of alumina-forming austenitic stainless steel","authors":"Guoshuai Chen ,&nbsp;Shang Du ,&nbsp;Lingzhi Chen ,&nbsp;Weiwei Cong ,&nbsp;Zhangjian Zhou","doi":"10.1016/j.jnucmat.2025.155785","DOIUrl":"10.1016/j.jnucmat.2025.155785","url":null,"abstract":"<div><div>The impact of substituting Mn for Ni on the microstructure and tensile properties of alumina-forming austenitic (AFA) stainless steel was systematically studied. The findings revealed that the addition of 4 wt. % Mn in place of 2 wt. % Ni could inhibit the precipitation of the B2-NiAl phase but increase the aspect ratio of the B2-NiAl particles and promote the precipitation of the Laves phase. The addition of Mn also promoted the formation of coincidence site lattice (CSL) grain boundaries and Goss texture, therefore beneficial for improving the mechanical properties. After aging at 700 °C, the room temperature (RT) ultimate tensile strength (UTS) and elongation of Mn-added AFA steel significantly improved to 1037.5 MPa and 34.53 %, respectively, compared to the Mn-free AFA steel, which exhibited a UTS of 848.35 MPa and elongation of 26.4 %. Notably, when tested at 700 °C, the elongation of Mn-added steel reached 60.5 %, nearly double that of Mn-free steel (36.5 %), while maintaining similar strength.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155785"},"PeriodicalIF":2.8,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768658","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 generation mechanism and performance degradation of CdZnTe radiation detectors in neutron radiation field","authors":"Qinzeng Hu ,&nbsp;Lingyan Xu ,&nbsp;Zhixin Tan ,&nbsp;Ming Hao ,&nbsp;Lu Liang ,&nbsp;Yingming Wang ,&nbsp;Zhentao Qin ,&nbsp;Lixiang Lian ,&nbsp;Chongqi Liu ,&nbsp;Yanyan Lei ,&nbsp;Wei Zheng ,&nbsp;Wanqi Jie","doi":"10.1016/j.jnucmat.2025.155781","DOIUrl":"10.1016/j.jnucmat.2025.155781","url":null,"abstract":"<div><div>Semiconductor radiation detectors used in nuclear power plants and other environments are inevitably exposed to neutron, γ-ray and other high-energy radiation, which can damage the crystal structure of semiconductors and thus degrade the detector performance. Here, we investigate the effects of neutron irradiation on the microstructure, photoelectric and radiation detection performance of CdZnTe detectors. Low-temperature photoluminescence (PL) spectra show that the dislocation related defect concentration in the irradiated crystals increases with increasing fluence. The infrared (IR) transmittance of the irradiated crystal decreases compared with that of the unirradiated crystal, which also indicates an increase in the dislocation density. The presence of stacking faults, stacking fault dipoles and dislocation locks in the irradiated CdZnTe crystals has been revealed by transmission electron microscopy (TEM). The energy resolution of γ-ray from ²⁴¹Am@100 V is degraded from 5.86 % before irradiation to 10.72 % after irradiation at 5.6 × 10¹⁰ n/cm². In addition, the mobility-lifetime product of electron (μτ)_e in CdZnTe detectors is reduced from 4.8 × 10^-3 cm²/V before irradiation to 7.02 × 10^-4 cm²/V after irradiation at 5.6 × 10¹⁰ n/cm². I-V test show that the barrier height of the CdZnTe detector decreases with the increase of neutron irradiation fluence, leading to a decrease in resistivity. Time-of-flight (TOF) tests demonstrate that the electron mobility after irradiation decreases with increasing irradiation fluence. Notably, the maximum neutron fluence used in this study is 3.9 × 10¹¹ n/cm², at which the CdZnTe radiation detector is not completely damaged. This study mainly investigates the radiation damage mechanism, induced defect characteristics and performance degradation of CdZnTe crystals by neutron irradiation, aiming to provide theoretical guidance for improving the radiation-resistant properties of detectors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155781"},"PeriodicalIF":2.8,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143734527","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}