Journal of Nuclear Materials最新文献_第7页

The influence of different vacancies on Zr(0001)/SiC close-packed interface performance: A first-principles study 不同空位对 Zr(0001)/SiC 紧密堆积界面性能的影响：第一原理研究

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-21 DOI: 10.1016/j.jnucmat.2024.155461

Yaojun Li , Sirui Liu , Hailin Zhai , Yuexia Wang , Yan Zhao , Xianfeng Ma

{"title":"The influence of different vacancies on Zr(0001)/SiC close-packed interface performance: A first-principles study","authors":"Yaojun Li , Sirui Liu , Hailin Zhai , Yuexia Wang , Yan Zhao , Xianfeng Ma","doi":"10.1016/j.jnucmat.2024.155461","DOIUrl":"10.1016/j.jnucmat.2024.155461","url":null,"abstract":"<div><div>The first-principles were employed to investigate the structure, adhesion, and tensile properties of the Zr(0001)/SiC close-packed interface with different vacancies. From the perspective of vacancy formation energy, SiC coating is beneficial for enhancing the irradiation resistance of Zr cladding. When vacancies are present, except for the Zr2 and Zr3 vacancies, introducing other vacancies reduces the stability of Zr/SiC interfaces. The C-terminated interface is more stable than the Si-terminated interface. Through electronic structure analysis, vacancies at the interface primarily reduce the bonds between Zr and SiC, decreasing the interface stability. Vacancies on the side of SiC indirectly alter the strength or quantity of covalent (bonding or anti-bonding) and ionic bonds at the interface, thus intricately lowering the interface stability. In tensile tests, the cleavage of all interfaces with vacancies still occurs on the side of Zr. Vacancies on the SiC side partly lead to increased electrons between Zr1-Zr2 or Zr2-Zr3, strengthening the metallic bonds and enhancing the interface's ideal strength and ductility. The present study offers a novel perspective from the standpoint of bonding mechanisms, providing good insights into the effects of different vacancies on the performance of Zr/SiC interfaces.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155461"},"PeriodicalIF":2.8,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571504","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Development of ZrSiO4-borosilicate glass-ceramics for immobilization of simulated tetravalent actinides 开发用于固定模拟四价锕系元素的 ZrSiO4 硼硅酸盐玻璃陶瓷

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-20 DOI: 10.1016/j.jnucmat.2024.155472

Hui Dan, Yihang Li, Bingbing Bao, Jiajing Li, Jiyuan Guo, Yi Ding

引用次数: 0

An investigation on the surface properties of B4C for advancing its nuclear applications 研究 B4C 的表面特性以促进其核应用

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-18 DOI: 10.1016/j.jnucmat.2024.155465

Jun Zhou, Nancy Lai Mun Wong, Jianwei Chai, Shijie Wang

{"title":"An investigation on the surface properties of B4C for advancing its nuclear applications","authors":"Jun Zhou, Nancy Lai Mun Wong, Jianwei Chai, Shijie Wang","doi":"10.1016/j.jnucmat.2024.155465","DOIUrl":"10.1016/j.jnucmat.2024.155465","url":null,"abstract":"<div><div>B<sub>4</sub>C is an important material in diverse nuclear applications. However, a systematic examination of its surface properties is still missing. In this work, we employ first-principles simulations to investigate the energetic stability of 16 distinct slab models representing (001), (100), (101), (110), and (111) surfaces, which are constructed by minimizing dangling bonds. Our results show that C-terminated (001) surface exhibits significantly greater stability than other surfaces under both the carbon and boron-rich conditions. Besides, we also study the defect formation energies on the C-terminated (001) surface and compare them with the cases in bulk. The high formation energies of the defects suggest a low likelihood of their occurrence on this surface, despite their formation energies being lower compared to bulk cases. Furthermore, mid-gap surface states are revealed for the top atomic layers of the C-terminated (001) surface, which are deduced at the deeper layers, and the band structures of the middle layers of this slab recover to the bulk band gap. These surface mid-gap states allow electron excitation from the valence band to these states, resulting in a reduced optical band gap compared to the bulk band gap of B<sub>4</sub>C. This provides a plausible explanation for the significantly smaller band gap observed in experiments compared to the larger gap predicted by theoretical models. Our study not only sheds light on the surface properties of B<sub>4</sub>C but also lays the groundwork for advancing this material for more advanced nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155465"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537320","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 irradiation-enhanced interstitial diffusion in over-pressurizing fission gas bubbles in UO2 辐照增强间隙扩散在二氧化铀过压裂变气泡中的作用

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-18 DOI: 10.1016/j.jnucmat.2024.155452

M.W.D. Cooper, C. Matthews, D.A. Andersson

{"title":"The role of irradiation-enhanced interstitial diffusion in over-pressurizing fission gas bubbles in UO2","authors":"M.W.D. Cooper, C. Matthews, D.A. Andersson","doi":"10.1016/j.jnucmat.2024.155452","DOIUrl":"10.1016/j.jnucmat.2024.155452","url":null,"abstract":"<div><div>Fission gas bubbles in UO<sub>2</sub> nuclear fuel have been observed to exhibit pressures in excess of the equilibrium bubble pressure; however, the cause of bubble over-pressurization has not yet been demonstrated. The mechanical interaction between a bubble and the surrounding matrix or grain boundary depends on the internal pressure of the bubble and local stress state, such that over-pressurized bubbles are thought to be responsible for fragmentation and pulverization, when exposed to a temperature ramp. Here, we investigate the role of U interstitials, produced through irradiation, in over-pressurizing bubbles by using a combined molecular dynamics (MD) and cluster dynamics approach. Firstly, the energies for the capture of interstitials and vacancies by bubbles have been determined from MD as a function of the ratio of gas atoms to vacancies that make up the bubble. Secondly, these reaction energies have been implemented in the cluster dynamics code Centipede to predict bubble over-pressurization as a function of temperature for typical fission rates. It was found that there is a transition from low pressure bubbles (at high temperatures) to high pressure bubbles (at lower temperatures). The cause of this behavior was shown to be the creation of irradiation-induced interstitials that are highly mobile relative to vacancies at low temperature; whereas, vacancies are sufficiently mobile at high temperatures to limit bubble pressures. This result supports the hypothesis that over-pressurized bubbles form during steady-state operation and that this behavior is highly sensitive to the local pellet temperature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155452"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142652837","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Corrosion-mediated production of uranium(III) chloride from metallic uranium in molten LiCl–KCl salt contained within a stainless-steel crucible 在不锈钢坩埚中的熔融 LiCl-KCl 盐腐蚀介导的金属铀氯化铀(III)生成过程

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-18 DOI: 10.1016/j.jnucmat.2024.155463

Eun-Young Choi , Seungwoo Paek , Taehyoung Kim , In-Ho Jung , Seol Kim , Sang-Eun Bae , Jae Soo Ryu

{"title":"Corrosion-mediated production of uranium(III) chloride from metallic uranium in molten LiCl–KCl salt contained within a stainless-steel crucible","authors":"Eun-Young Choi , Seungwoo Paek , Taehyoung Kim , In-Ho Jung , Seol Kim , Sang-Eun Bae , Jae Soo Ryu","doi":"10.1016/j.jnucmat.2024.155463","DOIUrl":"10.1016/j.jnucmat.2024.155463","url":null,"abstract":"<div><div>Uranium (III) chloride (UCl<sub>3</sub>) is a crucial component of a potent nuclear recycling technology—pyroprocessing—and next-generation molten salt reactors. It is usually synthesized by reacting metallic uranium with chlorinating agents (e.g., CdCl<sub>2</sub> and PbCl<sub>2</sub>) in molten chloride salts. In this study, we report the unexpected formation of UCl<sub>3</sub> from metallic simulated fuel (simfuel) immersed in impure molten LiCl–KCl salt (in the presence of a small amount of residual H<sub>2</sub>O) in a stainless-steel (SS) crucible, without a chlorinating agent. We investigated various factors influencing UCl<sub>3</sub> formation, including fuel type (metallic simfuel, pure U, oxide simfuel, or no fuel), crucible material (SS or alumina), salt composition (LiCl–KCl or LiCl), temperature (773 K or 923 K), and contact between fuel and SS crucible. UCl<sub>3</sub> only formed when metallic fuels (simfuel or pure U) were immersed in molten salt in the SS crucible, with higher concentrations at elevated temperatures. Oxide fuels did not produce UCl<sub>3</sub>, nor did contact with the crucible affect formation. Our findings suggest that impurities, particularly moisture in the salt, corroded the SS crucible, releasing iron and chromium chlorides that reacted with metallic U to form UCl<sub>3</sub>. UCl<sub>3</sub> formation was more pronounced in LiCl–KCl than in LiCl, and thermodynamic calculations helped establish the mechanism.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155463"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537328","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 study on the fretting corrosion of 316L in static lead-bismuth eutectic (LBE): The role of slip amplitude and normal force on damage mechanism at 350 °C 静态铅铋共晶（LBE）中 316L 的摩擦腐蚀研究：350 °C 时滑移幅度和法向力对损坏机制的作用

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-18 DOI: 10.1016/j.jnucmat.2024.155466

Hui Chen , Wenjie Pei , Shengzan Zhang , Wei Tan , Guorui Zhu

{"title":"A study on the fretting corrosion of 316L in static lead-bismuth eutectic (LBE): The role of slip amplitude and normal force on damage mechanism at 350 °C","authors":"Hui Chen , Wenjie Pei , Shengzan Zhang , Wei Tan , Guorui Zhu","doi":"10.1016/j.jnucmat.2024.155466","DOIUrl":"10.1016/j.jnucmat.2024.155466","url":null,"abstract":"<div><div>Fretting corrosion of stainless steel in the LBE affects the safety of lead-cooled fast reactors. Slip amplitude and normal load are the main mechanical factors affecting fretting wear behavior. Thus, the damage mechanism of 316L stainless steel at 350 °C LBE influenced by slip amplitude and normal load was investigated by jointly utilizing multiple characterization methods. The results indicate that the normal load and slip amplitude essentially affect the tangential stress and relative sliding value in the contact area, leading to different slip regions and damage mechanisms. In the mixed slip region, the damage mechanism is adhesion and delamination cracks. The increase in tangential stress leads to decrease in relative sliding. The thick wear debris layer attached to the worn surface can protect the substrate from being attacked by the LBE. In the gross slip region, the damage mechanism is abrasive wear and dissolution corrosion. The increase in relative sliding causes more damage and Ni dissolution, leading to the transformation from austenite to ferrite and internal strain, making the substrate more susceptible to damage and increasing the risk of liquid metal embrittlement (LME) of austenitic stainless steel at 350 °C. Accordingly, a model for different damage mechanisms was proposed. These results can provide important information on the fretting damage related to the LBE environment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155466"},"PeriodicalIF":2.8,"publicationDate":"2024-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142553738","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 new approach to investigate secondary hydriding phenomenon on M5Framatome clads under high–temperature LOCA conditions 研究高温 LOCA 条件下 M5Framatome 堆焊体二次水化现象的新方法

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-17 DOI: 10.1016/j.jnucmat.2024.155462

A.M. Kpemou , S. Guilbert , J. Desquines , T. Taurines , M.C. Baietto , B. Normand , J. Soulacroix , A. Ambard , F. Bourlier

{"title":"A new approach to investigate secondary hydriding phenomenon on M5Framatome clads under high–temperature LOCA conditions","authors":"A.M. Kpemou , S. Guilbert , J. Desquines , T. Taurines , M.C. Baietto , B. Normand , J. Soulacroix , A. Ambard , F. Bourlier","doi":"10.1016/j.jnucmat.2024.155462","DOIUrl":"10.1016/j.jnucmat.2024.155462","url":null,"abstract":"<div><div>The focus of this study is about a new experimental approach for a separate effects study of the secondary hydriding phenomenon under LOCA conditions. Many nuclear institutes perform semi–integrals tests to study the cladding behaviour during a LOCA transient. Those tests combined several phenomena and performing a detailed analysis of the secondary hydriding phenomenon using these tests can be challenging. A dedicated experimental protocol aiming at simulating secondary hydriding has been set up. Separate effects tests (SETs) were then carried out using this protocol to study the effects of both oxidation duration and temperature, on the hydrogen absorption during the oxidation stage of the LOCA transient on M5<sub>Framatome</sub><span><span><sup>1</sup></span></span> cladding. The effects of gap size were also investigated. Metallographic analysis has been used to characterise the M5<sub>Framatome</sub> clad metallurgical transformation after the high–temperature (HT) oxidation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155462"},"PeriodicalIF":2.8,"publicationDate":"2024-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537326","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

Demystify radiation-enhanced hydrogen isotope diffusion in Fe-Ni-Cr austenitic stainless steels 揭开铁-镍-铬奥氏体不锈钢中辐射强化氢同位素扩散的神秘面纱

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-16 DOI: 10.1016/j.jnucmat.2024.155460

X.W. Zhou, M.E. Foster

{"title":"Demystify radiation-enhanced hydrogen isotope diffusion in Fe-Ni-Cr austenitic stainless steels","authors":"X.W. Zhou, M.E. Foster","doi":"10.1016/j.jnucmat.2024.155460","DOIUrl":"10.1016/j.jnucmat.2024.155460","url":null,"abstract":"<div><div>Understanding and containing hydrogen isotope diffusion is crucial for many nuclear applications. In situ experiments have consistently shown that radiation significantly enhances isotope diffusion in austenitic stainless steels. Despite extensive research, the mechanism behind this phenomenon remains elusive, as most radiation-induced defects (e.g., vacancies, dislocations, and grain boundaries) typically trap hydrogen, thereby slowing diffusion. While grain boundaries may increase in-plane diffusivity and interstitials may enhance diffusion due to material swelling, these effects are relatively minor. Utilizing an Fe-Ni-Cr-H interatomic potential for stainless steels, we conducted extensive molecular dynamics simulations to investigate the origins of radiation-enhanced diffusion. Our findings reveal that when a system is resolidified, mimicking defects created by radiation displacements, the resulting structure contains a mixture of phases, boundaries, and dislocation networks. This defective structure significantly increases hydrogen diffusivity, enhancing it by approximately 1.7 times at 900 K. These results suggest that the complex defect structures formed during radiation displacements are the primary drivers of the observed diffusion enhancement, providing valuable insights into the mechanisms underlying radiation-enhanced diffusion in nuclear materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155460"},"PeriodicalIF":2.8,"publicationDate":"2024-10-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537319","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

Effect of ultrasonic field on the friction and oxidation characteristics of FeCrAl coatings 超声波场对铁铬铝涂层摩擦和氧化特性的影响

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-13 DOI: 10.1016/j.jnucmat.2024.155457

Changhao Liu, Xiufang Cui, Guo Jin, Meng Qi, Jiaxin Zhao, Di Wu, Xin Wen

引用次数: 0

Enhancing corrosion resistance of T91 F/M steel in liquid lead-bismuth (LBE) by slurry FeAl coating 通过铁铝浆涂层提高 T91 F/M 钢在液态铅铋 (LBE) 中的耐腐蚀性能

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-13 DOI: 10.1016/j.jnucmat.2024.155458

Wen Wang , Liujie Yang , Hongchen Qian , Zhaoguang Zhu , Guangjian Zhu , Jibo Tan , Jinyang Huang , Jintao Lu , Wenjun Kuang

引用次数: 0