Journal of Nuclear Materials最新文献_第2页

Insights into stress corrosion crack growth of 316NG heat-affected zone in simulated PWR primary water 模拟压水堆原生水316NG热影响区应力腐蚀裂纹扩展研究

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-09 DOI: 10.1016/j.jnucmat.2025.156028

Jun Xiao, Ting Xiao, Yu Qi Zheng, Yu Xiang Zhao, Qi Xu

{"title":"Insights into stress corrosion crack growth of 316NG heat-affected zone in simulated PWR primary water","authors":"Jun Xiao, Ting Xiao, Yu Qi Zheng, Yu Xiang Zhao, Qi Xu","doi":"10.1016/j.jnucmat.2025.156028","DOIUrl":"10.1016/j.jnucmat.2025.156028","url":null,"abstract":"<div><div>The SCC propagation behavior in the HAZ of 316NG remains insufficiently characterized. This study systematically evaluates SCC growth in HAZ specimens under simulated PWR primary water conditions (325 °C, 1200 mg/L B, 2 mg/L Li, both hydrogenated [30 mL (STP)/kg H₂] and oxygenated [0.5 ppm O₂] environments) at constant <em>K</em> = 30 MPa·m¹/². Compact tension (CT) specimens that contained artificial cracks positioned 1 mm and 4 mm from the weld fusion line were employed to asses SCC growth in the HAZ, where the crack propagation region extended to a normalized distance of 0.41–0.45 from the inner wall surface. Surprisingly, the HAZ exhibited crack growth rates marginally lower than or essentially comparable to those of the parent metal in both oxygenated and hydrogenated water environments. This suggests that compressive residual stresses in the SCC propagation region may mitigate the crack growth acceleration typically induced by strain hardening (up to 20 % hardness increase relative to the parent metal). High-resolution transmission electron microscopy (HRTEM) analysis demonstrated nickel (Ni) enrichment at grain boundaries ahead of advancing crack tips, attributed to rapid iron (Fe) diffusion along grain boundaries toward the crack tip and selective oxidation of Fe and chromium (Cr) at the crack tip. The oxide film formed a distinct bilayer structure, with an outer Fe₃O₄ magnetite layer and an inner FeCr₂O₄ spinel layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156028"},"PeriodicalIF":2.8,"publicationDate":"2025-07-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144632914","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

Computation of sink strengths in complex microstructures 复杂微结构中汇强度的计算

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-08 DOI: 10.1016/j.jnucmat.2025.156021

T. Jourdan , G. Adjanor

引用次数: 0

Differential effects of Si/dpa ratio on the evolution of dislocation loops in Al-2.64 Mg alloy during in-situ irradiation 原位辐照过程中Si/dpa比对Al-2.64 Mg合金位错环演化的差异影响

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-08 DOI: 10.1016/j.jnucmat.2025.156016

Shuo Cong , Ming Cao , Zuojiang Wang , Ziqi Cao , Ling Li , Yizhong Yang , Zhengyu Liu , Xujia Wang , Guang Ran

{"title":"Differential effects of Si/dpa ratio on the evolution of dislocation loops in Al-2.64 Mg alloy during in-situ irradiation","authors":"Shuo Cong , Ming Cao , Zuojiang Wang , Ziqi Cao , Ling Li , Yizhong Yang , Zhengyu Liu , Xujia Wang , Guang Ran","doi":"10.1016/j.jnucmat.2025.156016","DOIUrl":"10.1016/j.jnucmat.2025.156016","url":null,"abstract":"<div><div>Transmutated Si element is an important issue in the neutron irradiation of Al alloys. The Si production per dpa has been noticed to vary with the neutron energy spectrum, but the effect of Si/dpa ratio on the dislocation loop evolution has been rarely studied. In this work, the effect of Si/dpa ratio in Al-2.64 Mg alloy is in-situ quantitatively investigated by Si<sup>+</sup> irradiation at 100 °C with 1665, 468, and 57 appm/dpa. The higher the Si/dpa ratio, the smaller the loop size, but the higher the loop density in the matrix, which can be attributed to the high Si/dpa ratio leading to more lattice distortion and nucleation sites. Dislocation loops are preferentially formed around Al<sub>9</sub>Fe<sub>2-x</sub>Ni<sub>x</sub> precipitates due to the lattice mismatch at the interface. This phenomenon is more obvious at the low Si/dpa ratio. While Mg<sub>2</sub>Si precipitates have less effect on the formation of loops regardless of the Si/dpa ratio, because Al/Mg<sub>2</sub>Si interfaces are more likely to form a semi-coherent orientation and only result in mild lattice mismatch. But the type of dislocation loops is not affected by the lattice distortion and mismatch. In general, the Si atoms existed in Al lattice can significantly affect the dislocation loop evolution depending on the Si/dpa ratio. The current Si<sup>+</sup> irradiation results provide valuable data for understanding the transmutation behavior of Al alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156016"},"PeriodicalIF":2.8,"publicationDate":"2025-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654645","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

Temperature effects on radiation damage in HCP-zirconium: A molecular dynamics study using a fine-tuned machine-learned potential 温度对hcp -锆辐射损伤的影响：使用微调机器学习电位的分子动力学研究

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-08 DOI: 10.1016/j.jnucmat.2025.156025

Xinfang Jia, Yu Bao, Shuo Cao, Ye Su, Ping Qian

{"title":"Temperature effects on radiation damage in HCP-zirconium: A molecular dynamics study using a fine-tuned machine-learned potential","authors":"Xinfang Jia, Yu Bao, Shuo Cao, Ye Su, Ping Qian","doi":"10.1016/j.jnucmat.2025.156025","DOIUrl":"10.1016/j.jnucmat.2025.156025","url":null,"abstract":"<div><div>Zirconium has become an indispensable material in nuclear reactors due to its excellent corrosion resistance and low neutron absorption cross section. In this study, we fine-tune an efficient machine-learned interatomic potential for the pure Zr system. This potential demonstrates better performance in predicting defect properties, allowing us to conduct a comprehensive investigation of primary radiation damage through molecular dynamics simulations. We explore the threshold displacement energies of Zr at different temperatures and find that the variation in <span><math><msub><mrow><mi>E</mi></mrow><mrow><mi>d</mi></mrow></msub></math></span> with temperature is closely related to defect migration and recombination process. A series of cascade simulations with primary knock-on atom energies from 1 to 40 keV is conducted to investigate the effect of temperature on defect generation and clustering behavior in pure Zr. The results show that 10 keV serves as a critical value which large vacancy cluster begins to emerge, revealing distinct regimes of energy-dependence for defects. At low primary knock-on atom (PKA) energies, higher temperatures reduce both the number of steady defects and small-sized clusters. In contrast, at high PKA energies, the steady Frenkel pairs number <span><math><msub><mrow><mi>N</mi></mrow><mrow><mtext>s</mtext></mrow></msub></math></span> increases gradually with temperature, along with a greater occurrence of small-sized defect cluster. Moreover, it is likely to form large-sized defect clusters at lower temperature. Our findings provide critical insights into the irradiation damage mechanisms in Zr, offering theoretical guidance for the optimization of its radiation resistance.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156025"},"PeriodicalIF":2.8,"publicationDate":"2025-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144597292","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

In-reactor stress relaxation of 3-point bending CW-316 samples: Experiments and modelling 三点弯曲CW-316试样的堆内应力松弛：实验与模型

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-07 DOI: 10.1016/j.jnucmat.2025.155990

Guillaume Zumpicchiat , Benoit Tanguy , Pierre Gavoille , Valérie Vandenberghe , Julien Vidal

{"title":"In-reactor stress relaxation of 3-point bending CW-316 samples: Experiments and modelling","authors":"Guillaume Zumpicchiat , Benoit Tanguy , Pierre Gavoille , Valérie Vandenberghe , Julien Vidal","doi":"10.1016/j.jnucmat.2025.155990","DOIUrl":"10.1016/j.jnucmat.2025.155990","url":null,"abstract":"<div><div>3-point bending in-pile stress relaxation tests of cold-worked 316 stainless steel have been characterised in the mixed-spectrum OSIRIS reactor up to 7 dpa at 330°C. Relaxation kinetics were determined based on out of pile interphase measurements. The experimental design allows to study initial prescribed stress levels from 100 MPa up to 600 MPa and to investigate transient (primary) and steady-state (secondary) stress relaxation. Relaxation kinetics were found to be unaffected by initial prescribed stress but to depend on the cold work level of the samples. No incubation dose was observed, even for samples loaded at low initial stresses. Experimental results were interpreted using FE modelling to take into account the non-uniform stress distribution existing in 3-point bending samples loaded in the elasto-plastic domain. The linearly stress-dependent creep law, consisting of two terms representing transient and steady-state creep regimes, allows for a very good description of experimental relaxation kinetics all over the investigated stress range. The primary creep amplitude parameter was found to be dependent on the cold work level, which is consistent with literature. A steady state creep coefficient of 0.96 10<sup>−6</sup> MPa<sup>−1</sup>.dpa<sup>−1</sup> consistent with the values reported from fast reactors was obtained independently on the cold work level. Also it was shown that the stress relaxation ratio can be directly determined based on deflection relaxation ratio despite the non-uniform stress state and residual stresses into the samples.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 155990"},"PeriodicalIF":2.8,"publicationDate":"2025-07-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144581113","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

Flow-Accelerated Corrosion and Wall Thinning Mechanism of a Dissimilar Metal Welded Joint in the Secondary Loop Piping System of Pressurized Heavy Water Reactors 重水堆二次回路管道系统异种金属焊接接头流动加速腐蚀及壁薄机理

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-06 DOI: 10.1016/j.jnucmat.2025.156026

Yuling You , Liang Zhao , Jiamei Wang , Hui Zheng , Kai Chen , Lefu Zhang

{"title":"Flow-Accelerated Corrosion and Wall Thinning Mechanism of a Dissimilar Metal Welded Joint in the Secondary Loop Piping System of Pressurized Heavy Water Reactors","authors":"Yuling You , Liang Zhao , Jiamei Wang , Hui Zheng , Kai Chen , Lefu Zhang","doi":"10.1016/j.jnucmat.2025.156026","DOIUrl":"10.1016/j.jnucmat.2025.156026","url":null,"abstract":"<div><div>This study investigates the flow-accelerated corrosion (FAC) failure in a dissimilar metal weld joint (DMWJ) of a decommissioned secondary loop piping elbow from a nuclear plant after 20 years of operation in China. Comprehensive macroscopic and microscopic failure characterizations were conducted to elucidate the wall thinning mechanism. High-temperature electrochemical galvanic corrosion tests and computational fluid dynamics (CFD) simulations were performed to validate the proposed thinning mechanism. The results reveal that the SA106B carbon steel (CS) piping near the weld joint exhibits a typical scallop-shaped two-phase FAC morphology. Notably, the maximum wall thinning, reaching 1.404 mm, was observed in the SA106B heat-affected zone (HAZ) adjacent to the fusion boundary. The galvanic corrosion tests and CFD calculations confirm that the wall thinning is governed by a combination of galvanic corrosion, driven by the potential difference between CS and stainless steel (SS), and significant localized flow turbulence induced by weld reinforcement. The localized flow turbulence increases wall shear stress (WSS), liquid film thickness, and liquid droplet impingement (LDI), contributing to both erosive and corrosive effects, ultimately enhancing wall thinning. Special attention should be given to evaluating the thinning at the weld fusion line during non-destructive failure examinations of piping systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156026"},"PeriodicalIF":2.8,"publicationDate":"2025-07-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144613385","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

Utilizing a machine-learned potential to explore enhanced radiation tolerance in the MoNbTaVW high-entropy alloy 利用机器学习潜力探索MoNbTaVW高熵合金的增强辐射耐受性

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-05 DOI: 10.1016/j.jnucmat.2025.156004

Jiahui Liu , Jesper Byggmästar , Zheyong Fan , Bing Bai , Ping Qian , Yanjing Su

{"title":"Utilizing a machine-learned potential to explore enhanced radiation tolerance in the MoNbTaVW high-entropy alloy","authors":"Jiahui Liu , Jesper Byggmästar , Zheyong Fan , Bing Bai , Ping Qian , Yanjing Su","doi":"10.1016/j.jnucmat.2025.156004","DOIUrl":"10.1016/j.jnucmat.2025.156004","url":null,"abstract":"<div><div>High-entropy alloys (HEAs) based on tungsten (W) have emerged as promising candidates for plasma-facing components in future fusion reactors, owing to their excellent irradiation resistance. To achieve physically realistic descriptions of primary radiation damage in such multi-component materials, we propose extended damage models and trained an efficient machine-learned interatomic potential for the MoNbTaVW quinary system. From cascade simulations at primary knock-on atom (PKA) energies of 1–150 keV, we fitted an extended arc-dpa model for quantifying radiation damage in MoNbTaVW. Furthermore, we performed 50 cascade simulations at the recoil energy of 150 keV with 27.648 million atoms to investigate the effect of PKA types (Mo, Nb, Ta, V, W). The results show that subcascade splitting effectively suppresses interstitial cluster formation, which is a key mechanism for enhancing radiation resistance in HEAs. Our findings provide valuable insights into the radiation resistance mechanisms in refractory body-centered cubic alloys and highlight the potential of machine learning approaches in radiation damage research.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156004"},"PeriodicalIF":2.8,"publicationDate":"2025-07-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144581114","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

3D microscopic X-ray fluorescence and diffraction analysis of a fuel-cladding interface of a highly irradiated boiling water reactor fuel rod 高辐照沸水堆燃料棒燃料包壳界面三维显微x射线荧光和衍射分析

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-04 DOI: 10.1016/j.jnucmat.2025.156020

C. Schneider , M.  G. Makowska , J. Bertsch

{"title":"3D microscopic X-ray fluorescence and diffraction analysis of a fuel-cladding interface of a highly irradiated boiling water reactor fuel rod","authors":"C. Schneider , M.  G. Makowska , J. Bertsch","doi":"10.1016/j.jnucmat.2025.156020","DOIUrl":"10.1016/j.jnucmat.2025.156020","url":null,"abstract":"<div><div>During irradiation of nuclear fuel in light water reactors (LWRs), the fuel cladding creeps onto the uranium oxide fuel pellets due to the high temperature and pressure in the reactor core. With increasing burnup, the uranium fission leads to a surplus of oxygen that creates an oxide layer at the inner surface of the cladding which causes a fuel-cladding bonding. This bonding layer, that incorporates fission products, can be important for the pellet-cladding mechanical interaction. The present study focuses on the fuel-cladding interface (FCI) region of a high burnup spent fuel irradiated in a Swiss Boiling Water Reactor (61 GWd/tU). A microstructural characterization was carried out using a combined microbeam X-ray fluorescence (µXRF) and X-ray diffraction (µXRD) experiment in tomographic mode. All the measurements were carried out at the microXAS beamline of the Swiss Light Source synchrotron, at the Paul Scherrer Institute, PSI. A small micrometer-scale sample of a size of approximately 34 µm x 34 µm x 18 µm of the region of interest at the FCI was prepared using the focused ion beam (FIB) technique and then analyzed. The µXRF maps, obtained with high spatial resolution, were used to correlate the chemical composition with the crystalline phases identified using XRD. The results show that tetragonal zirconia is the main phase present in the bond layer between fuel and cladding, although it is a metastable form under the temperature and pressure conditions of a LWR. Lattice parameters evolve through the zirconia layer, with smaller parameters at the Zr/ZrO<sub>2</sub> interface and larger ones at the ZrO<sub>2</sub>/UO<sub>2</sub> interface. This may indicate phase stabilization by stresses and/or fission products. XRD analysis allowed evaluating the lattice parameters of the identified phases, among others for UO<sub>2</sub> near the cladding, which was 5.474 Å. In addition, XRF analyses revealed the presence of Kr bubbles within the zirconia layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156020"},"PeriodicalIF":2.8,"publicationDate":"2025-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144633238","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 probable radiation-induced precipitate hcp-Re in heavy ion irradiated Mo-42Re alloy 重离子辐照Mo-42Re合金中可能的辐射诱发沉淀hcp-Re

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-04 DOI: 10.1016/j.jnucmat.2025.156024

Kelong Lu , Yongjin Wu , Shijian Lin , Daqing Yuan , Ke Li , Ping Fan , Hailiang Ma , Hang Xu , Hao Yang , Hongtao Huang

引用次数: 0

Influence of Cu/W semi-coherent interfacial structure on helium behavior: insights from atomistic simulations Cu/W半相干界面结构对氦行为的影响：来自原子模拟的见解

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-07-03 DOI: 10.1016/j.jnucmat.2025.156022

Wen Chen, Qing Hou, Tun Chen, Jie-Chao Cui, Min Li, Bao-Qin Fu

引用次数: 0