Journal of Nuclear Materials最新文献_第2页

Dissolution behavior of U3O8, FeUO4, and UO2-Zr-stainless steel system samples generated in an oxidative atmosphere in the presence of malonic acid

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155561

Ryutaro Tonna , Takayuki Sasaki , Yoshihiro Okamoto , Taishi Kobayashi

{"title":"Dissolution behavior of U3O8, FeUO4, and UO2-Zr-stainless steel system samples generated in an oxidative atmosphere in the presence of malonic acid","authors":"Ryutaro Tonna , Takayuki Sasaki , Yoshihiro Okamoto , Taishi Kobayashi","doi":"10.1016/j.jnucmat.2024.155561","DOIUrl":"10.1016/j.jnucmat.2024.155561","url":null,"abstract":"<div><div>Understanding the dissolution behavior of the UO<sub>2</sub>-Zr-stainless steel (SS) system in the presence of organic acids is crucial for the safe decommissioning of the Fukushima Daiichi Nuclear Power Plants, as well as for the future treatment toward deep geological disposal of fuel debris. The dissolution behavior of UO<sub>2</sub>-Zr-SS samples heated under oxidative conditions, along with the single uranium phases formed in the samples (U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub>)<sub>,</sub> was investigated through static leaching tests using malonic acid. Malonic acid promoted the dissolution of both U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub> solid phases owing to complex formation. The uranium concentrations of U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub> increased with malonic acid concentration and matched at steady state. In the absence of malonic acid, the U concentration of FeUO<sub>4</sub> was less than its solubility because of uranium adsorption on an iron hydrolysis solid phase. The uranium concentration after long immersion of the UO<sub>2</sub>-Zr-SS system samples could be explained by the behavior of U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub>. In the absence of malonic acid, U concentration decreased owing to adsorption onto the iron solid phase, similar to the behavior of FeUO<sub>4</sub>. In contrast, in the presence of malonic acid, U concentration was consistent with that observed for U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub>.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155561"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171602","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

A model for trapping and re-solution regarding intra-granular bubbles in UO2, linked to atomic-scale simulations

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155562

M. Vergani , M.W.D. Cooper , L. Noirot

{"title":"A model for trapping and re-solution regarding intra-granular bubbles in UO2, linked to atomic-scale simulations","authors":"M. Vergani , M.W.D. Cooper , L. Noirot","doi":"10.1016/j.jnucmat.2024.155562","DOIUrl":"10.1016/j.jnucmat.2024.155562","url":null,"abstract":"<div><div>In the literature, a clear definition of the irradiation re-solution frequency of gas from bubbles in the UO<sub>2</sub> fuel is absent. Moreover, for intra-granular bubbles, a detailed calculation of the cumulated displaced gas quantities in function of the distance from the radius of the bubble after a re-solution event has never been published. The assessment of these two elements is very useful if we want to increase the adherence of fission gas release codes to our present knowledge of the behavior of fission gases. Hence, we suggest to link the definition of the re-solution frequency to atomic-scale simulations. Furthermore, we present the cumulated displaced gas quantities obtained from Molecular Dynamics calculations, from which we have derived a re-solution profile that can be exploited to better consider the irradiation re-solution phenomenon inside Fission Gas Release codes. On top of that, we have built a new trapping/re-solution model for intra-granular bubbles linked to Molecular Dynamics simulations that can be easily incorporated into Fission Gas Release codes. We also check that the model is properly built through the comparison of the new model against a reference.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155562"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171603","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

Characterization of the microstructure of yttrium hydride under proton irradiation

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155586

Stephen Taller , Fabian Naab , Takaaki Koyanagi , Timothy Lach

{"title":"Characterization of the microstructure of yttrium hydride under proton irradiation","authors":"Stephen Taller , Fabian Naab , Takaaki Koyanagi , Timothy Lach","doi":"10.1016/j.jnucmat.2024.155586","DOIUrl":"10.1016/j.jnucmat.2024.155586","url":null,"abstract":"<div><div>High moderation per unit volume solid moderator materials like yttrium hydride (YH<sub>x</sub>) are necessary for compact nuclear microreactors. However, the phase stability and hydrogen transport processes of YH<em><sub>x</sub></em> under high-temperature irradiation are largely unknown. Proton irradiation was conducted on YH<em><sub>x</sub></em> at 300 °C and 580 °C to 0.2 dpa using 1 MeV or 2 MeV protons in a high-vacuum environment. The hydrogen concentration was determined before and after irradiation using elastic recoil detection analysis, and microstructural evolution was examined via post-irradiation scanning transmission electron microscopy and Raman spectroscopy. Dislocation loops and cavities were observed in all conditions; their distribution was correlated with the bombarding proton energy and ion irradiation temperature. This work revealed that hydrogen retention is proportional to the formation of traps for hydrogen gas atoms and identified pathways for hydrogen release. The relative contributions of bulk or fast diffusion paths, such as grain boundaries, delamination boundaries, and stacking faults are discussed; the primary mechanisms of hydrogen loss are likely based on diffusion, ruling out artefacts of the experimental design. The study suggests proton irradiation may be a strong surrogate to study hydrogen transport in hydride moderator materials under irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155586"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154923","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

Fracture behavior and grain boundary cohesion of alumina scales formed on ion-irradiated FeCrAl-ODS alloy

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155663

Hao Yu , Diancheng Geng , Yasuyuki Ogino , Naoko Oono-Hori , Koji Inoue , Sosuke Kondo , Ryuta Kasada , Shigeharu Ukai

{"title":"Fracture behavior and grain boundary cohesion of alumina scales formed on ion-irradiated FeCrAl-ODS alloy","authors":"Hao Yu , Diancheng Geng , Yasuyuki Ogino , Naoko Oono-Hori , Koji Inoue , Sosuke Kondo , Ryuta Kasada , Shigeharu Ukai","doi":"10.1016/j.jnucmat.2025.155663","DOIUrl":"10.1016/j.jnucmat.2025.155663","url":null,"abstract":"<div><div>The design of FeCrAl ferritic oxide dispersion strengthened (ODS) alloys is based on the formation of a stable alumina scale, which is expected to protect the alloys from extreme heat and corrosion in nuclear applications. To ensure reliable alumina protection in nuclear environment, it is indispensable to concern the radiation tolerance of the alumina scales formed on the FeCrAl ODS alloys. The present study investigates the effect of Fe ions irradiation on fracture modes and grain boundary cohesion of the alumina scales in conjunction with nano-impact tests and micro-double notch shear (DNS) compression tests. Pre-oxidation was carried out in air at 1000 °C to form an α-alumina layer on the surface of Fe-15Cr-7Al-0.5Y<sub>2</sub>O<sub>3</sub>–0.4Zr (wt.%) ferritic ODS alloy, followed by 6.4 MeV Fe<sup>3+</sup> ion beam irradiation at 500 °C. Based on the microstructural characterization of the cross-sectional micrographs of nanoindentation imprints on the alumina scales, it was confirmed that the irradiation on the alumina scales resulted in significant intergranular fracture in nanoindentation, whereas the unirradiated alumina scales showed transgranular fracture. The elemental distribution around the alumina grain boundaries was elucidated with the aid of scanning transmission electron microscopy (STEM) and atom probe tomography (APT) observations, and obvious segregation of reactive elements (REs) and intergranular Ti/TiC precipitation were observed after irradiation, indicating the link between the microstructural evolution and the fracture behavior of the alumina scales. The detailed grain boundary cohesion of alumina scales before and after irradiation was accurately measured by the micro-DNS compression tests, and the results showed that the cohesion strength of the alumina decreased significantly after the Fe ions irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155663"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155303","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}

引用次数: 0

Effects of He+ energy and W temperature on the initial W fuzz growth under the fusion-relevant He+ irradiation

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155630

Weifeng Liu, Chunjie Niu, Weiyuan Ni, Dongping Liu

{"title":"Effects of He+ energy and W temperature on the initial W fuzz growth under the fusion-relevant He+ irradiation","authors":"Weifeng Liu, Chunjie Niu, Weiyuan Ni, Dongping Liu","doi":"10.1016/j.jnucmat.2025.155630","DOIUrl":"10.1016/j.jnucmat.2025.155630","url":null,"abstract":"<div><div>The growth of tungsten nanofuzz (W fuzz) induced by helium ions (He<sup>+</sup>) irradiation is a critical issue for fusion devices such as ITER. In our study, we investigated the behavior of tungsten (W) under helium ion (He<sup>+</sup>) irradiation, focusing on the conditions that promote the growth of tungsten fuzz (W fuzz). A comprehensive model was utilized to analyze the impact of varying He<sup>+</sup> energies and W temperatures on the W fuzz formation and growth. It was found that W fuzz was formed in a low-energy range from ∼10 eV to 200 eV and a higher-energy range of >5 keV. W fuzz growth was facilitated below the He-W sputtering threshold energy while sputtering erosion became a significant hindrance just above this threshold. At He<sup>+</sup> energies reaching hundreds of eV, W fuzz growth was entirely suppressed due to the enhanced sputtering erosion. However, at higher He<sup>+</sup> energies around 5 keV, the He<sup>+</sup> penetration reduced the impact of sputtering erosion, allowing He bubbles to grow in the deeper W layer. W temperature varying in the range of 1000 – 2000 K played a crucial role in forming He bubbles – induced tensile stress in the W surface layer, therefore affecting W fuzz growth which strongly depended on the He<sup>+</sup> energy. These insights provided a detailed understanding of the energy-temperature relevance for W fuzz formation, which was crucial for predicting the performance and lifetime of W components in future fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155630"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155344","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 microstructural evolution in irradiation hardening of Alloy 718 under low dose proton irradiation

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155644

Haipeng Zhu , Zhiming Zhang , Jianqiu Wang , Hongliang Ming , Zhiyuan Zhang , Yilan Jiang , Quanyao Ren , En-Hou Han

{"title":"The role of microstructural evolution in irradiation hardening of Alloy 718 under low dose proton irradiation","authors":"Haipeng Zhu , Zhiming Zhang , Jianqiu Wang , Hongliang Ming , Zhiyuan Zhang , Yilan Jiang , Quanyao Ren , En-Hou Han","doi":"10.1016/j.jnucmat.2025.155644","DOIUrl":"10.1016/j.jnucmat.2025.155644","url":null,"abstract":"<div><div>The microstructural evolution and hardness changes of Alloy 718 under low dose (≤0.09 dpa) proton irradiation were investigated. Low dose irradiation induced the formation of defect clusters and disordering of the γ″ phase, while no voids or radiation-induced segregation were observed. The density of defect clusters increased with increasing irradiation dose, but their size did not change significantly. The irradiation damage behavior varied among different γ″ variants, with those having their [001] direction parallel to the proton beam being the most resistant to disordering. The hardness of the irradiated material was simultaneously affected by the hardening due to irradiation defects and the softening caused by the disordering of the γ″ phase. The competition between these two factors led to a decrease in hardness at 0.02 dpa and 0.04 dpa compared to the unirradiated material, while at 0.09 dpa, the hardness increased above that of the unirradiated material.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155644"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155392","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

Cesium sorption on UO2: systematic experimental and model studies for surface sites characterization.

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155614

N. Rodríguez-Villagra , U. Alonso , J. Cobos

引用次数: 0

Microstructural and mechanical responses of 316H and weld metal under Helium irradiation at 550 °C

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155564

Dong Wang , Lianyong Xu , Lei Zhao , Yongdian Han

{"title":"Microstructural and mechanical responses of 316H and weld metal under Helium irradiation at 550 °C","authors":"Dong Wang , Lianyong Xu , Lei Zhao , Yongdian Han","doi":"10.1016/j.jnucmat.2024.155564","DOIUrl":"10.1016/j.jnucmat.2024.155564","url":null,"abstract":"<div><div>The irradiation microstructure and nanoindentation of 316H base metal and weld metal after He ions irradiation were systematically studied at 550 °C. The evolution of He bubbles and Frank loops were quantitatively characterized. Rate theory calculation demonstrated that the bias for interstitial atoms induced by high density of dislocation in 316H weld metal resulted in the low number density of Frank loops and larger sized He bubbles. Segregation of He bubbles in the <em>γ</em>/<em>δ</em> interface of 316H weld metal was also observed. The irradiation hardening in 316H weld metal was lower than that in the 316H base metal. Microstructure based calculation showed that the Frank loops dominated the irradiation hardening in 316H base metal. Whereas, in the 316H weld metal, Frank loops dominated the irradiation hardening under the low irradiation fluence, and He bubbles dominated the irradiation hardening under the high irradiation fluence. The irradiation hardening induced by Frank loops in 316H weld metal was obviously alleviated compared with that in 316H base metal. The insights into the microstructure evolution and irradiation hardening mechanism in irradiated 316H base metal and 316H weld metal can benefit the structural integrity assessment and optimization of austenitic steel in Gen-IV nuclear energy system.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155564"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170546","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

Kinetics and mechanism of oxidation of simulated spent nuclear fuel segment at 500°C

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155581

Yandong Sun, Zheng Wei, Ying Chen, Fang Liu, Taihong Yan, Tianchi Li, Zhongwei Yuan, Weifang Zheng

引用次数: 0

Microstructural stability and mechanical property of novel high-Si high-Cr reduced activation ferritic/martensitic steels at high temperatures

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155602

Qianfu Pan , Sen Ge , Chao Sun , Gaixia Wang , Yu Wu , Xiaoe Xu , Huiqun Liu

{"title":"Microstructural stability and mechanical property of novel high-Si high-Cr reduced activation ferritic/martensitic steels at high temperatures","authors":"Qianfu Pan , Sen Ge , Chao Sun , Gaixia Wang , Yu Wu , Xiaoe Xu , Huiqun Liu","doi":"10.1016/j.jnucmat.2024.155602","DOIUrl":"10.1016/j.jnucmat.2024.155602","url":null,"abstract":"<div><div>The present work investigated the microstructural stability and mechanical property of four novel high-Si and high Cr reduced activation ferritic/martensitic steels at elevated temperature. Alloy plate samples were normalized at 1373 K for 1 h, tempered at 1023 K for 1 h, and then aged at 873 K for 1000, 2000, and 3000 h In the tempered state, M<sub>23</sub>C<sub>6</sub> precipitates were distributed along grain and lath boundaries, while MX precipitates were uniformly dispersed in the matrix containing different amounts of ferrites, which was similar to with the calculated result. The microstructure of the designed alloys exhibited high-thermal stability even after 3000 h aging, with the martensitic grain size and ferrite content nearly unchanged. However, M<sub>23</sub>C<sub>6</sub> were coarsened with increasing the aging time. Additionally, with increasing W content, the coarsening rate significantly decreased. After aging for 1000 h, the designed alloys precipitated needle-like Laves phases with a faster coarsening rate, and the size and volume fraction increased with W and Si content. While VN precipitates exhibited significantly higher stability, maintaining a constant particle size (60 ∼ 80 nm) even after aging for 3000 h, which is attributed to variations in the diffusion coefficients of elements. The designed alloys exhibited high yield strength (488 ∼ 548 MPa at room temperature) in the 3000 h-aged state, surpassing that of commercial EP823 (462 MPa), where the strengthening mechanisms were also discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155602"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154914","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