Journal of Nuclear Materials最新文献_第4页

Comparative studies of the long-term corrosion behavior of Zr-Sn-Fe-Cr-Ni alloys in pure water at 360 °C/18.6 MPa with high and low dissolved oxygen content 360 °C/18.6 MPa 溶解氧含量高低条件下纯水中 Zr-Sn-Fe-Cr-Ni 合金长期腐蚀行为的比较研究

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-24 DOI: 10.1016/j.jnucmat.2024.155481

Aijia Lei , Xun Dai , Yufeng Du , Jingjing Liao , Ruiju Deng , Jiangtao Xu , Xuefei Huang

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Mechanical and high-temperature steam oxidation properties of Cr coatings deposited via high-power impulse magnetron sputtering 通过高功率脉冲磁控溅射沉积的铬涂层的机械和高温蒸汽氧化特性

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-24 DOI: 10.1016/j.jnucmat.2024.155482

Ding Chen , Wei Dai , Daoxuan Liang , Qimin Wang , Jun Yan

{"title":"Mechanical and high-temperature steam oxidation properties of Cr coatings deposited via high-power impulse magnetron sputtering","authors":"Ding Chen , Wei Dai , Daoxuan Liang , Qimin Wang , Jun Yan","doi":"10.1016/j.jnucmat.2024.155482","DOIUrl":"10.1016/j.jnucmat.2024.155482","url":null,"abstract":"<div><div>Applying protective coatings to Zr alloy cladding surfaces is one of the better methods to design fuel tolerant materials. In this study, the surface of a Zr-4 alloy was coated with Cr using high-power impulse magnetron sputtering. Furthermore, the mechanisms by which bias voltages affect the mechanical characteristics, resistance to high-temperature steam oxidation, and coating structure were elucidated. The coating exhibits a strong (200) weave structure with coarse grains at a bias voltage of -100 V. With increasing bias, the energy of deposited particles increases, grains continue to grow, (200) preferential growth orientation disappears, and the coating exhibits a (110) crystal orientation. The growth structure of the coating first shows a tendency to be dense and then loose. For the Cr coating with a (200) crystal orientation, a dense oxide layer is preferentially formed after oxidation, which can effectively block the internal diffusion of O. With increasing oxidation time, coarse Cr grains can effectively block the external diffusion of Zr. Furthermore, the Cr coating exhibiting a (110) crystal orientation was severely oxidized after oxidation, resulting in the formation of cracks at the film base; this accelerated the outward diffusion of Zr.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155482"},"PeriodicalIF":2.8,"publicationDate":"2024-10-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142537329","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}

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DFT simulations of the self-healing behavior of a W〈110〉/W〈112〉 grain boundary in the presence of coexisting point defects 共存点缺陷下 W〈110〉/W〈112〉晶界自愈行为的 DFT 模拟

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-23 DOI: 10.1016/j.jnucmat.2024.155471

J. Suárez-Recio , D. Fernández-Pello , M.A. Cerdeira , C. González , R. Gonzalez-Arrabal , R. Iglesias

{"title":"DFT simulations of the self-healing behavior of a W〈110〉/W〈112〉 grain boundary in the presence of coexisting point defects","authors":"J. Suárez-Recio , D. Fernández-Pello , M.A. Cerdeira , C. González , R. Gonzalez-Arrabal , R. Iglesias","doi":"10.1016/j.jnucmat.2024.155471","DOIUrl":"10.1016/j.jnucmat.2024.155471","url":null,"abstract":"<div><div>Light impurity atoms (LIAs), such as hydrogen and helium, tend to aggregate at pre-existing intrinsic point defects. This aggregation leads to detrimental effects, particularly in environments such as those foreseen in nuclear fusion reactors. There, such impurities would be ubiquitous, resulting in unacceptable material behavior that would unqualify the material as a Plasma Facing Material (PFM). One option to delay the degradation in performance is the use of nanostructured tungsten (NW), showing a large density of grain boundaries (GBs). Although we have already addressed the behavior of a single LIA in a GB, in this work we present the combined synergistic effects of the simultaneous presence of multiple LIAs, vacancies and Self-Interstitial Atoms (SIA) at semicoherent W/W interfaces using ab initio methods. Our results reveal a complex and interesting process in the competition between LIAs and SIAs. When the number of SIAs is low, He appears to hinder their recombination with vacancies, therefore casting doubts on the self-healing provided by NW. However, in the presence of larger numbers of SIAs, their mutual repulsion leads to the opposite behavior. Thus, a thorough thermodynamic assessment in which the evolution of the system may be tracked emerges as the crucial subsequent step in these investigations.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155471"},"PeriodicalIF":2.8,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142560922","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}

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Demonstration of industrially-fabricated plutonium disposition MOX 工业化生产的钚处置 MOX 演示

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-23 DOI: 10.1016/j.jnucmat.2024.155477

Claire L. Corkhill , Latham T. Haigh , Lewis R. Blackburn , Luke T. Townsend , Daniel J. Bailey , Lucy M. Mottram , Amber R. Mason , Max R. Cole , Thierry Gervais , Genevieve Kerboul

{"title":"Demonstration of industrially-fabricated plutonium disposition MOX","authors":"Claire L. Corkhill , Latham T. Haigh , Lewis R. Blackburn , Luke T. Townsend , Daniel J. Bailey , Lucy M. Mottram , Amber R. Mason , Max R. Cole , Thierry Gervais , Genevieve Kerboul","doi":"10.1016/j.jnucmat.2024.155477","DOIUrl":"10.1016/j.jnucmat.2024.155477","url":null,"abstract":"<div><div>The safe and secure management of civil separated plutonium is a UK government and NDA priority. One potential solution to address this considers the manufacture of a modified version of mixed oxide (MOX) fuel, comprising PuO<sub>2</sub> dispersed within a UO<sub>2</sub> matrix and doped with a suitable neutron absorbing element to maintain criticality control. As an initial step to understand whether an industrially-relevant, proven MOX fuel fabrication process could offer a potential route to the production of a Pu-disposition matrix based on MOX, a series of Gd-doped UO<sub>2</sub> pellets were prepared by Orano at the CDA workshop of the MELOX facility in France. Characterisation was performed to quantify the density, morphology (grain size and porosity), Gd distribution and Gd incorporation mechanism. It was found that the materials produced were highly reproducible and similar in density and morphology, irrespective of the variables investigated, and similar to unirradiated UOX and MOX fuel. Gd was distributed in a similar manner to the distribution of PuO<sub>2</sub> in unirradiated MIMAS (MIcronisation of a MASter Blend) MOX fuel and evidence for the existence of a solid solution between Gd<sub>2</sub>O<sub>3</sub> and UO<sub>2</sub> was ascertained, which could be viewed as favourable from a GDF post-closure criticality control perspective. The source of the powder had the greatest effect on the final characteristics of the Pu-disposition MOX pellets, due to sintering reactivity; however, these differences were minor. These results are a promising step towards the full-scale manufacture of ceramics suitable for the immobilisation and disposition of separated PuO<sub>2</sub> in a GDF, should policy dictate.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155477"},"PeriodicalIF":2.8,"publicationDate":"2024-10-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571505","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}

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Microstructure and corrosion property evolution of a surface-nanostructured 15–15Ti austenitic steel during immersion in liquid LBE at 550 °C 表面纳米结构 15-15Ti 奥氏体钢在 550 °C 液态 LBE 中浸泡期间的微观结构和腐蚀性能演变

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-22 DOI: 10.1016/j.jnucmat.2024.155475

Da Wang , Weiqian Zhuo , Sirui Liu , Changquan Xiao , Wenjian Zhu , Bihan Sun , Xianfeng Ma , Ganfeng Yuan , Yulin Sun

{"title":"Microstructure and corrosion property evolution of a surface-nanostructured 15–15Ti austenitic steel during immersion in liquid LBE at 550 °C","authors":"Da Wang , Weiqian Zhuo , Sirui Liu , Changquan Xiao , Wenjian Zhu , Bihan Sun , Xianfeng Ma , Ganfeng Yuan , Yulin Sun","doi":"10.1016/j.jnucmat.2024.155475","DOIUrl":"10.1016/j.jnucmat.2024.155475","url":null,"abstract":"<div><div>This study investigated the compatibility of surface-nanostructured 15–15Ti austenitic steel in 550 °C LBE with an oxygen concentration of 5 × 10<sup>−7</sup> wt.% for various exposure durations (759, 1638, 2404, and 3012 h). The results demonstrate that the grain size was reduced from 33.50 μm to the nano-scale after shot-peening (SP), achieving 17.62, 15.44, and 14.25 nm under SP pressures of 0.06, 0.15 and 0.25 MPa, respectively. The untreated steel experienced severe oxidation and dissolution corrosion, whereas the surface-nanostructured steel exhibited only mild oxidation and was resistant to dissolution corrosion. The enhanced corrosion resistance of surface-nanostructured steel is attributed to the higher protectiveness of the Cr-rich spinel layer and the less defective Ni-rich layer beneath it. Recrystallization occurred exclusively in the Ni-rich region, while the deformed steel underwent recovery during exposure. The thickness of the recrystallization layer was 2.9 μm at 759 h, increased to 8 μm at 1638 h, and remained stable thereafter. The size of recrystallized grains in SP-samples processed under pressure of 0.06 MPa and 0.15 MPa was approximately 2.92 μm, whereas it was about 1.32 μm for 0.25 MPa processed sample.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155475"},"PeriodicalIF":2.8,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142553739","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}

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Pore structure evolution of A3–3 matrix graphite during heat treatment A3-3 基质石墨在热处理过程中的孔隙结构演变

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-22 DOI: 10.1016/j.jnucmat.2024.155474

Xi Tong, Xiangwen Zhou, Kaihong Zhang, Huixun Gao, Shouchi Zhang, Bing Liu, Yaping Tang

{"title":"Pore structure evolution of A3–3 matrix graphite during heat treatment","authors":"Xi Tong, Xiangwen Zhou, Kaihong Zhang, Huixun Gao, Shouchi Zhang, Bing Liu, Yaping Tang","doi":"10.1016/j.jnucmat.2024.155474","DOIUrl":"10.1016/j.jnucmat.2024.155474","url":null,"abstract":"<div><div>Matrix graphite (MG), a key component of fuel elements for high-temperature gas-cooled reactors (HTRs), has a profound effect on the comprehensive performance and service safety of fuel elements. A3–3 MG was selected as the matrix material for the pebble fuel elements of the 10 MW experimental high-temperature gas-cooled reactor (HTR-10) and the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) in China. During the preparation process of A3–3 MG, the green MG pebble must undergo two-stage heat treatment, namely carbonization and purification, to obtain excellent comprehensive properties for safe service. However, the porosity of A3–3 MG and its change during heat treatment remains unclear. Herein, the pore structure evolution through three different stages of A3–3 MG - the green, carbonized and purified samples- were tested using the gas adsorption method, mercury intrusion porosimetry and X-ray computed tomography (X-CT). The green sample had the smallest pore diameter and a uniform pore size distribution. The pore structure of the carbonized sample was the most developed, with the most micropores, mesopores and macropores. The molecular-sized micropores were produced due to the pyrogenic decomposition of the resin binder. Purification led to a decrease in pore diameter, together with a slight increase in closed pores and a decrease in pore connectivity due to pore merging and conversion. Two- and three-dimensional (2D and 3D) pore structure models were established by X-CT scan. The variation in pore size and shape, different types of pores as well as the pore conversion during the heat treatment process of A3–3 MG were observed. In this work, the porosity evolution of A3–3 MG was studied in detail, and references and strategies were provided for optimizing the preparation process and performance of pebble fuel elements.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155474"},"PeriodicalIF":2.8,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142579110","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}

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Corrigendum to “Experimental assessment of thermodynamic stability and nucleation of NiO in liquid lead-bismuth eutectic for MYRRHA” [Journal of Nuclear Materials 603 (2025) 155404] 对 "用于 MYRRHA 的液态铅铋共晶中氧化镍的热力学稳定性和成核的实验评估"[《核材料学报》603 (2025) 155404]的更正

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-22 DOI: 10.1016/j.jnucmat.2024.155467

Aleksandr Tsybanev , Alessandro Marino , Jun Lim , Kristof Gladinez , Nele Moelans

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Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles 辐照增强银在 TRISO 粒子中扩散的多尺度机理建模

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-21 DOI: 10.1016/j.jnucmat.2024.155464

Pierre-Clément A. Simon, Jia-Hong Ke, Chao Jiang, Larry K. Aagesen, Wen Jiang

{"title":"Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles","authors":"Pierre-Clément A. Simon, Jia-Hong Ke, Chao Jiang, Larry K. Aagesen, Wen Jiang","doi":"10.1016/j.jnucmat.2024.155464","DOIUrl":"10.1016/j.jnucmat.2024.155464","url":null,"abstract":"<div><div>Tristructural isotropic (TRISO) particles are under consideration for use in several proposed advanced nuclear reactor concepts. The silicon carbide (SiC) layer in TRISO acts as a barrier to prevent the release of the fission products. However, despite remarkable retention, silver (Ag) release has been observed from intact particles, which requires investigation since the Ag isotope (<span><math><msup><mrow></mrow><mrow><mn>110</mn><mi>m</mi></mrow></msup></math></span>Ag) has a long half-life. Previous work focused on developing a multiscale, mechanistic model for Ag diffusion accounting for temperature and microstructure effect and has been successfully validated. In this work, we expand the previous model to account for irradiation-enhanced Ag diffusivity in SiC and improve its accuracy over a wider grain size and temperature ranges relevant for advanced reactor conditions. A temperature, grain size, and flux dependent diffusivity is therefore derived using the mesoscale code MARMOT and implemented in the fuel performance code BISON. The irradiation-enhanced Ag diffusivity in SiC is compared against experimental data and validated using BISON against Ag release measurements from the Advanced Gas Reactor Fuel Development and Qualification Program (AGR-1 and AGR-2). Herein, we quantify the impact of SiC grain size, irradiation, and temperature on Ag release. In agreement with previous studies, we find accounting for SiC grain size improves agreement between BISON predictions and experimental observations for most cases. We also find that accounting for irradiation improves agreement for cases where Ag release was underestimated, but the impact was less significant than accounting for microstructure.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155464"},"PeriodicalIF":2.8,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142553740","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}

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In-situ study on the differential evolution of He bubbles in the multilayer oxide of Fe9Cr1.5W0.4Si F/M steel corroded in lead-bismuth eutectic 铅铋共晶中腐蚀的 Fe9Cr1.5W0.4Si F/M 钢多层氧化物中 He 气泡差异演化的原位研究

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2024-10-21 DOI: 10.1016/j.jnucmat.2024.155473

Dewang Cui , Shuo Cong , Ziqi Cao , Fan Yuan , Guang Ran

{"title":"In-situ study on the differential evolution of He bubbles in the multilayer oxide of Fe9Cr1.5W0.4Si F/M steel corroded in lead-bismuth eutectic","authors":"Dewang Cui , Shuo Cong , Ziqi Cao , Fan Yuan , Guang Ran","doi":"10.1016/j.jnucmat.2024.155473","DOIUrl":"10.1016/j.jnucmat.2024.155473","url":null,"abstract":"<div><div>The combined effects of corrosion and irradiation on nuclear components have been an important but not yet fully revealed topic. Here, the irradiation behavior of the oxide scale formed on F/M steel after lead-bismuth corrosion was in-situ investigated during He<sup>+</sup> irradiation. The results showed that the oxide scale included a Fe<sub>3</sub>O<sub>4</sub> outer oxide layer, a nanograin Fe(Fe<sub>x</sub>Cr<sub>2-x</sub>)O<sub>4</sub> spinel inner oxide layer, and an internal oxide layer. He bubbles formed in Fe<sub>3</sub>O<sub>4</sub>, Fe-Cr spinel and F/M steel were polygon, irregular elongated pores and small spheres, respectively. These differences were attributed to variations in defect generation, migration, and corrosion-induced crystal defects in different oxides. Numerous corrosion-induced nanograin boundaries and vacancies in Fe-Cr spinel exhibited more effective absorption of irradiation-induced defects. Moreover, rhombic perfect dislocation loops were detected in Fe<sub>3</sub>O<sub>4</sub> at the late stage of irradiation, their relative positional relationship with He bubbles indicated a potential interaction between bubbles and loops.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155473"},"PeriodicalIF":2.8,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142554390","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}

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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