Journal of Nuclear Materials最新文献

{"title":"Corrigendum to “Pore structure development of oxidized nuclear graphite” [J. Nucl. Mater. 601 (2024) 155342]","authors":"Rong-Cheng Li, Qing Huang, Xiang-Lei Yao, He Zhou, Xing-Tai Zhou","doi":"10.1016/j.jnucmat.2024.155530","DOIUrl":"10.1016/j.jnucmat.2024.155530","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155530"},"PeriodicalIF":2.8,"publicationDate":"2024-11-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142745585","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":"Investigation of ion irradiation effects on mineral analogues of concrete aggregates","authors":"Zehui Qi ,&nbsp;Xin Chen ,&nbsp;Arnaud Bouissonnié ,&nbsp;Elena Tajuelo Rodriguez ,&nbsp;Yann Le Pape ,&nbsp;Miguel L. Crespillo ,&nbsp;Gaurav Sant ,&nbsp;Steven John Zinkle","doi":"10.1016/j.jnucmat.2024.155539","DOIUrl":"10.1016/j.jnucmat.2024.155539","url":null,"abstract":"<div><div>Irradiation can cause prominent damage to reactor concrete aggregates leading to amorphization, strength and modulus decrease, radiation induced volume expansion (RIVE) and micro-cracking, which limits their long-term performance. To develop an improved understanding of irradiation effects in concrete, three mineral analogues of concrete aggregates (limestone, marble and quartzite) were irradiated by 5.5 MeV He ions and 13 MeV Ni ions to surface doses of 0.011 displacements per atom (dpa) and 0.23 dpa, respectively, at room temperature. The two different ion species allow irradiation spectrum effects (ionizing and displacive) to be examined. Irradiation induced cracks were observed in He irradiated limestone and marble, and Ni irradiated quartzite. Full amorphization was observed in Ni irradiated quartzite with 14.3 % RIVE, and ∼25 % hardness and modulus decrease, while almost no change was observed in He irradiated quartzite except 4.35 % RIVE, revealing a possible ionization enhanced diffusion effect for high energy light ions. Furthermore, partial amorphization was observed in Ni irradiated marble and limestone matrix with a 12 % hardness decrease in marble while no amorphization was observed for He irradiation with a 20 % hardness increase in limestone matrix. The role of knock-on damage and irradiation spectrum on amorphization, volumetric expansion and mechanical property changes are discussed. Moreover, the onset and critical doses for amorphization and RIVE in quartz are obtained for ion irradiations at room temperature. The dose dependence of RIVE exhibits a delay compared to the amorphization behavior. The superior irradiation resistance of calcite phase compared to quartz phase implies there could be advantages to using calcareous aggregates and lowering the usage of siliceous aggregates for concrete in nuclear power plants for extended operation beyond 60 years. However, other effects such as corrosion, aging and reactions during severe accidents should also be considered, and further investigations are needed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155539"},"PeriodicalIF":2.8,"publicationDate":"2024-11-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142745581","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":"Diffusion coefficients calculations of 110mAg in ZrC at very high temperature using machine-learning interatomic potential","authors":"Jae Joon Kim ,&nbsp;Eung-Seon Kim ,&nbsp;Hyun Woo Seong ,&nbsp;Ho Jin Ryu","doi":"10.1016/j.jnucmat.2024.155532","DOIUrl":"10.1016/j.jnucmat.2024.155532","url":null,"abstract":"<div><div>The ternary machine-learning interatomic potential for a Zr–C–Ag system was developed using first-principles calculations, moment tensor potential, and molecular dynamics simulations to calculate the diffusion coefficient of Ag in ZrC. The developed potential was utilized to investigate the vacancy formation energy, Ag substitutional energy, binding energy between Ag and vacancy, Ag interstitial energy, migration energy of Ag to an adjacent C vacancy in ZrC, and thermal expansion of ZrC. The results conformed to previously reported experimental and computational results, thereby validating the accuracy of the developed potential. The diffusion coefficients of Ag in ZrC_0.94 and ZrC_0.97 in the temperature range of 2800–3200 K were calculated using molecular dynamics simulations with the developed machine-learning interatomic potential. These calculation results can aid the safety analysis of radioactive ^110mAg release in nuclear thermal propulsion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155532"},"PeriodicalIF":2.8,"publicationDate":"2024-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142745583","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":"Combined effects of metallic dopants and nonmetallic impurities on interface cohesion in tungsten alloys by first-principles","authors":"Y.X. Zhang ,&nbsp;Y.G. Zhang ,&nbsp;Z.M. Xie ,&nbsp;X.Y. Li ,&nbsp;Y.C. Xu ,&nbsp;R. Liu ,&nbsp;C.S. Liu ,&nbsp;X.B. Wu","doi":"10.1016/j.jnucmat.2024.155536","DOIUrl":"10.1016/j.jnucmat.2024.155536","url":null,"abstract":"<div><div>The effect of element segregation on interface cohesion demonstrates significant potential in tailoring the mechanical performances of materials. In this study, we have investigated the impact of co-segregation of transition metal (Re, Zr and Ti) and non-metallic impurity elements (O and C) on the cohesion properties of two typical interfaces, W/HfC phase boundary (PB) and Σ5(310) grain boundary (GB) in W alloys, using first-principles calculations. Our findings reveal that O atom exhibits comparable segregation tendency at both the PB and GB interfaces, but the PB has a stronger resistance to O-embrittlement than the GB. C atom preferentially segregates at the GB and enhances the interface cohesion. In addition, Re atoms tend to segregate at both the interfaces and enhance the interface cohesion. Co-segregation of Zr/Ti and O atoms at the interface leads to a reduction in impurity O concentration within the W matrix, and further decreases the interface cohesion. In contrast, C atom mitigates the GB embrittlement induced by Zr/Ti atom owing to the formation of W-C bonds. This work deepens the understanding of how the co-segregation of alloying and non-metallic impurity elements affects the interface properties, offering theoretical guidance for optimizing the mechanical performance of W-based materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155536"},"PeriodicalIF":2.8,"publicationDate":"2024-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142745584","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":"Evaluation of the thermal conductivity of chromium coatings for accident tolerant fuels using thermoreflectance techniques","authors":"Yuzhou Wang,&nbsp;Yaoyang Zhang,&nbsp;Hailin Zhai,&nbsp;Jishen Jiang,&nbsp;Yibo Zhang,&nbsp;Qiang Zhang,&nbsp;Xianfeng Ma","doi":"10.1016/j.jnucmat.2024.155535","DOIUrl":"10.1016/j.jnucmat.2024.155535","url":null,"abstract":"<div><div>The deployment of Cr coatings as accident tolerant fuel cladding inside nuclear reactors necessitates a thorough understanding of their performance, including mechanical, chemical, irradiation, and thermal properties. While significant research has delved into the former, the thermal transport properties of Cr coatings, which dictates fuel temperature and impacts fuel safety, have been relatively underexplored. This work presents an investigation into the thermal conductivity of 12 µm-thick Cr coatings fabricated by multi-arc ion plating employing high-resolution frequency and spatial domain thermoreflectance techniques. Vacuum annealing was utilized to control the morphology of coatings to explore the impact of grain size. The results indicate that both as-deposited and annealed coatings, despite possessing distinct grain morphologies, exhibit thermal conductivities similar to their bulk counterparts. The influence of boundaries on thermal transport is anticipated to become prominent as the critical dimension diminishes below 0.5 μm. Future investigation is required to address the potential impact of adverse reactions, such as oxygen ingress and irradiation, on the thermal properties of Cr coatings to ensure safe and efficient operation in nuclear reactor environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155535"},"PeriodicalIF":2.8,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142705994","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":"Non-electric mass transfer between stainless steel 316H and glassy carbon in NaF-KF-UF4 salt","authors":"Jaewoo Park,&nbsp;Jinsuo Zhang","doi":"10.1016/j.jnucmat.2024.155534","DOIUrl":"10.1016/j.jnucmat.2024.155534","url":null,"abstract":"<div><div>This study focuses on non-electric mass transfer between stainless steel 316H (SS316H) and carbon materials in molten NaF-KF-UF₄ salt. To simulate the molten-salt flow and accelerate the mass transfer, an SS316H specimen was rotated with a tangential speed of 2 m/s in the molten salt contained in a glassy carbon crucible at 1073 K for 120 h. The concentrations of Cr, Fe, Ni, and Mn in the salt were measured as a function of time during the test. The depletion of Cr was observed near the surface of the post-test SS316H specimen, and Fe-concentrated layers were formed on the specimen. (Cr, Fe)₇C₃ layers, Cr-metal particles, and dendritic structures concentrated with Fe and Cr were observed on the post-test crucible, indicating the non-electric mass transfer between SS316H and glassy carbon.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155534"},"PeriodicalIF":2.8,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142748572","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":"Grain boundary self-diffusion and point defect interactions in α-U via molecular dynamics","authors":"Khadija Mahbuba ,&nbsp;Benjamin Beeler ,&nbsp;Andrea Jokisaari","doi":"10.1016/j.jnucmat.2024.155521","DOIUrl":"10.1016/j.jnucmat.2024.155521","url":null,"abstract":"<div><div>Though metallic U-Zr fuel has been used in nuclear reactors since the 1960s, many of its fundamental and thermodynamic properties are still unknown. The <em>α</em>-U phase, which has a highly anisotropic crystal structure and physical properties, is present in U-Zr fuel. The character and behavior of <em>α</em>-U grain boundaries will strongly impact fuel thermophysical performance under irradiation. We study the interaction of point defects with grain boundaries, diffusion along grain boundaries, and the predicted diffusional creep behavior of <em>α</em>-U via molecular dynamics. We calculate the segregation energy of vacancies and interstitials to grain boundaries and quantify the biased sink strength of the grain boundaries, and observe that this sink strength is not strongly dependent on the grain boundary orientation. We also find that grain boundary diffusivity is strongly dependent on the grain boundary energy and grain boundary orientation. The presence of point defects within the grain boundary can induce diffusion in grain boundaries with low formation energies and can enhance diffusion in high-energy grain boundaries. We also find that diffusional creep of <em>α</em>-U at prototypical metallic fuel operation conditions is extremely high and could help explain observed metallic fuel swelling behaviors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155521"},"PeriodicalIF":2.8,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142705996","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":"Assessing ceramic powder quality by activated Sinterability Test: The case of UO2","authors":"Balakrishna Palanki","doi":"10.1016/j.jnucmat.2024.155531","DOIUrl":"10.1016/j.jnucmat.2024.155531","url":null,"abstract":"<div><div>Ceramics fail by brittle fracture due to flaws and affect process yield. The starting material is usually in powder form. UO₂ pellets are obtained by pressing powder, sintering and finish grinding. Large powder blends are usually accepted for pressing and sintering after evaluating a small representative powder sample by conducting a sinterability test under regular process conditions. On the other hand, this paper recommends activated sintering conditions, such as those achieved with additives or sintering atmosphere control. Many defects in ceramics have origins in the powder. For example, large hard agglomerates in the powder can cause packing difficulties in pressing. Defects that are not detected in normal sintering may be noticed more readily in activated sintering due to defect amplification. In sintering, open porosity ceases after reaching a density of ∼93 % TD. The residual closed porosity tends to shrink on further sintering. The temperature at which open porosity or permeability is lost shifts to a lower temperature in activated sintering. Yet, activated sintering is to be carried out at conventional high sintering temperature, to be able to amplify and expose pellet defects due to powder. Desintering is a result of large sized packing defects in the green body and premature loss of open porosity in the course of sintering. A descriptive model of desintering is suggested that takes into account powder specific surface area, sintering additive and atmosphere. There is no desintering when green microstructure is homogeneous with no density gradients and with uniformly distributed fine voids that shrink and close during sintering. A high-quality powder sample is one that results in high pelleting yield both in conventional and activated sintering. The low temperature sintering process for UO₂ manufacture that did not progress due to thermal stability concerns in nuclear reactor, may be revived to lower nuclear fuel costs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"604 ","pages":"Article 155531"},"PeriodicalIF":2.8,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142705999","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":"On the analysis of radiation-induced segregation at ion-irradiated grain boundaries","authors":"Daniele Fatto' Offidani ,&nbsp;Enrique Martinez ,&nbsp;Emmanuelle A. Marquis","doi":"10.1016/j.jnucmat.2024.155533","DOIUrl":"10.1016/j.jnucmat.2024.155533","url":null,"abstract":"<div><div>Radiation-induced segregation has been extensively studied in alloys irradiated with neutrons and charged particles (protons and ions). Unlike neutrons and protons, heavy ions have a very short penetration depth (typically a few micrometers), and for this reason, the free surface and end-of-range region can significantly affect radiation-induced microstructural changes. These effects are well-documented for void swelling and dislocation loops; however, they are much less so for radiation-induced segregation at grain boundaries. To address this gap, a model Fe–Ni–Cr alloy was irradiated using Fe ions and grain boundary chemistry was quantified as a function of depth and dose. The impacts of surface oxidation and grain boundary diffusion on radiation-induced segregation are discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155533"},"PeriodicalIF":2.8,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142745580","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 spatially-resolved model of neutron-irradiated tungsten coupling stochastic cluster dynamics and finite deformation plasticity","authors":"Sabyasachi Chatterjee ,&nbsp;Qianran Yu ,&nbsp;Yang Li ,&nbsp;Kenneth Roche ,&nbsp;Jaime Marian ,&nbsp;Giacomo Po","doi":"10.1016/j.jnucmat.2024.155526","DOIUrl":"10.1016/j.jnucmat.2024.155526","url":null,"abstract":"<div><div>Structural materials used in nuclear reactors face severe degradation in mechanical properties, such as hardening and embrittlement. At the microscopic scale, this occurs due to creation and accumulation of irradiation-induced defects and their interaction with system dislocations. Although techniques exist which can model evolution of irradiation defects, for instance kinetic transport theory-based models, their interaction with mechanical deformation of the bulk material has not been investigated extensively. In this work, we demonstrate a novel spatially-resolved multiscale coupling between microscopic irradiation defect evolution, modeled using Stochastic Cluster Dynamics (SCD) and macroscopic mechanical deformation modeled using a finite-deformation plasticity model. SCD is used to determine the statistically averaged defect cluster spacing, dependent on operating conditions such as irradiation dose and temperature. This acts as an initial condition that governs the critical resolved shear stress of dislocation glide in the macroscopic plasticity model. This framework is used to predict mechanical behavior in post-mortem test of irradiated Tungsten samples, which has found its importance as structural material used in nuclear reactors. The results obtained using the coupled approach are in good agreement with experimental data of uniaxial tension tests. The model is able to capture the effect of temperature and irradiation dose on the material hardening. Two methods are proposed to estimate hardness – using Tabor's Law relating uniaxial yield stress to hardness and from flat-punch simulations. The results are in reasonable agreement with hardness data from micro-indentation experiments of irradiated Tungsten samples. The model is also able to reveal microstructural details such as spatial variation in defect density and local stress.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155526"},"PeriodicalIF":2.8,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142721127","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}