Journal of Nuclear Materials最新文献

{"title":"Thermal conductivity of irradiated tetragonal lithium aluminate","authors":"Víctor H. Ortiz ,&nbsp;Weilin Jiang ,&nbsp;Andrew M. Casella ,&nbsp;David J. Senor ,&nbsp;Ram Devanathan ,&nbsp;S. Aria Hosseini ,&nbsp;P. Alex Greaney ,&nbsp;Richard B. Wilson","doi":"10.1016/j.jnucmat.2024.155585","DOIUrl":"10.1016/j.jnucmat.2024.155585","url":null,"abstract":"<div><div>The effect of 120 keV He⁺+<em>D</em>⁺-ion irradiation on the thermal conductivity of ceramic tetragonal γ-LiAlO₂ is studied with time-domain thermoreflectance (TDTR) at temperatures between 300 and 700 K. The thermal conductivity of single crystal γ-LiAlO₂ is 13.5 W/(m·K) at 300 K, and scales with temperature like 1/<em>T</em>. The thermal conductivity of unirradiated polycrystalline γ-LiAlO₂ is 7.4 W/(m·K). Irradiation at fluences of 1 ×, 5 ×, and 10 × 10¹⁶ ions/cm² decreases the thermal conductivity by ≈ 30 %, 80 %, and 90 %. The effect of irradiation is saturated at ion fluences of 10¹⁷ ions/cm². Irradiation decreases the temperature dependence of the thermal conductivity. For ion fluences larger than 10¹⁷ ions/cm², the thermal conductivity reaches a minimum value of ≈ 1 W/(m·K) that is independent of temperature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155585"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154922","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":"Some practical methodologies to assess the overall high temperature (one-sided) steam oxidation protectiveness of chromium-based coatings on a zirconium-based substrate, as Enhanced – Accident tolerant (Nuclear) fuels (E-ATF) claddings","authors":"Jean-Christophe Brachet,&nbsp;Thomas Guilbert,&nbsp;Stéphane Urvoy ,&nbsp;Elodie Rouesne,&nbsp;Marion Peyret,&nbsp;Thierry Vandenberghe,&nbsp;Cédric Prou,&nbsp;Thai Le Hong,&nbsp;Didier Hamon","doi":"10.1016/j.jnucmat.2025.155620","DOIUrl":"10.1016/j.jnucmat.2025.155620","url":null,"abstract":"<div><div>Among various Enhanced-Accident Tolerant nuclear fuel claddings studied worldwide, Cr-coated Zr-based claddings are largely considered as a reference with evolutionary design and resulting ability to improve the high-temperature (HT) oxidation resistance. Most of the numerous studies carried out so far on the HT oxidation behavior of coated claddings have focused on detailed localized <em>post-mortem</em> examinations. The present paper aims to suggest complementary and sufficiently simple experimental methodologies to evaluate the overall protectiveness of Cr-based coated materials upon HT steam oxidation. The different proposed approaches have been assessed by comparing their capacity to characterize the Cr-coating loss of protectiveness and the associated oxygen diffusion into the sub-coating metallic wall-clad thickness. The latter is then compared to the results obtained by “more direct” Electron-Probe-Micro-Analysis measurements. The overall oxygen ingress into the tested coated claddings, evaluated by different proposed approaches, can be correlated to the Post-Quenching strength of the claddings, using a novel “Equivalent Cladding Reacted” parameter - called “Equivalent Chromium Reacted” ECrR – recently developed at CEA.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155620"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155313","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":"The effect of nickel coatings on corrosion of nickel-chrome alloys in molten FLiBe determined using a phase-field model","authors":"Chaitanya Bhave,&nbsp;Michael R. Tonks","doi":"10.1016/j.jnucmat.2025.155616","DOIUrl":"10.1016/j.jnucmat.2025.155616","url":null,"abstract":"<div><div>An existing phase-field model (Bhave et al. (2023) <span><span>[19]</span></span>) is applied to investigate why there is so much variation reported in the literature on the effectiveness of pure Ni coatings in reducing corrosion by molten salt. We simulate the impact of Ni diffusion barrier coatings on the corrosion of Ni-Cr alloys by molten FLiBe using 2D simulations. We first compare the corrosion behavior in a Ni-20Cr alloy exposed to molten FLiBe at 700<!--> ^∘C with and without a pure Ni coating. The coating reduces the mass loss after 1000 hours by a factor of ten, consistent with experimental results from the literature. The model is then used with Latin hypercube sampling involving 100 simulations with different coating thicknesses and average alloy and coating grain sizes. As the coating grain size increases, the model predicts that the mass loss and corrosion depth into the alloy decreases. This is due to a decrease in the number of fast diffusion paths along the coating grain boundaries (GBs) for the Cr to reach the salt. As the alloy grain size increases, the model predicts that the mass loss decreases but the corrosion depth increases. This is because larger grain size creates less GB area for Cr depletion, increasing mass loss, but less GB area also allows the Cr depletion to penetrate further into the alloy. In addition, the model predicts that as the coating thickness increases, the mass loss rapidly decreases and the impact of both grain sizes also decreases. Thus, controlling the coating grain size is less important with thicker coatings.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155616"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155314","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":"First-principles molecular dynamics evaluation of the electroreduction kinetics of Nd from molten LiCl-KCl salt on different liquid metals","authors":"Xuemin Ding ,&nbsp;Yafei Wang ,&nbsp;Biao Wu ,&nbsp;Xinyu Zhang ,&nbsp;Shaoqiang Guo ,&nbsp;Weiqian Zhuo","doi":"10.1016/j.jnucmat.2025.155661","DOIUrl":"10.1016/j.jnucmat.2025.155661","url":null,"abstract":"<div><div>Electrorefining has been identified as a promising technique for recovering valuable actinides from spent nuclear fuels. However, the electrochemical separation of accumulated fission products in the electrorefining salt electrolyte is a remaining challenge that is important for the minimization of the radioactive waste volume. In this work, we employed first-principles molecular dynamics (FPMD) simulations to investigate the performance of five candidate liquid metal electrodes (i.e., Bi, Pb, Sb, Sn, Zn) in separating Nd (a major fission product) from LiCl-KCl molten salt. The equilibrium potentials and diffusion coefficients of Nd³⁺ in these liquid metal cathodes are calculated via FPMD simulations and validated with literature data. The radial distribution functions, bond-angle distribution functions, Voronoi polyhedron and binding energies corresponding to the local Nd structure are analyzed to interpret the strong atomic interaction between Nd and the liquid metals. Finally, the calculated properties are incorporated into an electrode kinetics model to evaluate the electrolysis performance of different liquid metal cathodes, demonstrating the prospect of FPMD-based approach to streamline the selection of efficient electrode materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155661"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155398","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":"Impact of dopants and leachants on modern UO2-based fuels alteration under final storage conditions: Single and joint effects","authors":"N. Rodríguez-Villagra ,&nbsp;S. Fernández-Carretero ,&nbsp;A. Milena-Pérez ,&nbsp;L.J. Bonales ,&nbsp;L. Gutiérrez ,&nbsp;J. Cobos ,&nbsp;H. Galán","doi":"10.1016/j.jnucmat.2025.155635","DOIUrl":"10.1016/j.jnucmat.2025.155635","url":null,"abstract":"<div><div>Doped-UO₂ fuels such as Cr- or Cr/Al-UO₂ (accident-tolerant fuels (ATF) or modern nuclear fuels) and Gd–UO₂ fuel (being Gd a burnable neutron absorber now in use in LWR fuels) need to be deeply studied not only relating to its advantages under normal and accident conditions in operation, but also its behavior under different repository conditions. After the geologic repository post-closure, once the spent nuclear fuel come into contact with groundwater after container failure, the release of some radionuclides will rely on the UO₂ matrix dissolution processes. The corrosion/dissolution behavior of doped UO₂ fuels, including ATF, in a deep geological repository is barely comprehended. The join influence of dopants and groundwater composition needs further enhance knowledge and understanding by filling gaps in the empirical databases. This study examines the impact of Cr, Cr/Al and Gd dopants on the corrosion of UO₂ fuel pellets in groundwater, based on the known benefits of adding certain soluble metal oxides to UO₂, depending on the nature of the doping element. Systematic dissolution experiments were conducted with Cr-, Cr/Al-, and Gd-doped UO₂ pellets in three aqueous media with varying pH and HCO₃⁻ concentrations. Groundwaters used with increasing the complexity of the system (<em>i.e.</em> contain many ionic species) were 20 mM NaClO₄ (pH 7.2), 19:1 mM NaHCO₃:NaCl (pH 8.9), and synthetic young cement water with Calcium (pH 13.5). The experiments revealed that the leachant attributes, particularly the combined effects of pH, redox conditions, and HCO₃⁻, had a more significant impact on uranium concentration and dissolution rates than the dopants themselves. No secondary uranium phases were observed on the surfaces of any post-leached samples. These findings contribute to the understanding of the combined effects of doping and aqueous composition on the dissolution behavior of modern nuclear fuels under long-term conditions anticipated in a deep geological repository.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155635"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155401","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":"Effects of long-term thermal aging on mechanical properties and microstructural evolution of 17–4 PH stainless steel in simulated thermal conditions for nuclear applications","authors":"Junho Lee ,&nbsp;Gitae Park ,&nbsp;Chang Young Oh ,&nbsp;Youngho Son ,&nbsp;Seunghyun Kim ,&nbsp;Chi Bum Bahn","doi":"10.1016/j.jnucmat.2025.155628","DOIUrl":"10.1016/j.jnucmat.2025.155628","url":null,"abstract":"<div><div>This study examined the effects of long-term thermal aging on the mechanical properties and microstructure of 17–4 PH stainless steel (SS) at temperatures from 300 °C to 400 °C for up to 12,000 h. Mechanical tests, including hardness, strength, and impact toughness tests, were conducted, along with microstructural analysis using transmission electron microscopy. The results indicated that aging at 400 °C leads to early embrittlement and a decrease in mechanical strength after 10,000 h of exposure, due mainly to spinodal decomposition and G-phase formation. At 350 °C, the formation of a G-phase was observed at the boundary between Cu precipitates and martensite matrix after 5,000 h, contributing significantly to the rapid decrease in toughness, but the hardness and mechanical strength were only minimally affected. In contrast, at 300 °C, the mechanical strength increased more gradually, with only spinodal decomposition influencing the mechanical behavior. In particular, slight softening was observed during the first 1,000 h at 300 °C and 350 °C because of carbon diffusion that promoted the growth of niobium and chromium carbides, weakening the martensitic matrix. This study highlights the significant role of microstructural evolution, particularly the relationship between the formation of the G-phase and impact toughness, in determining the long-term mechanical properties of 17–4 PH SS under prolonged thermal aging under simulated thermal conditions for nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155628"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154877","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":"Low-energy potential-induced helium trapping in nano-austenitic steels","authors":"Jun Hui ,&nbsp;Jiapeng Chen ,&nbsp;Min Liu ,&nbsp;Shuo Wang ,&nbsp;Biao Wang","doi":"10.1016/j.jnucmat.2025.155636","DOIUrl":"10.1016/j.jnucmat.2025.155636","url":null,"abstract":"<div><div>In austenitic steel for use in reactor cores, the selected solute elements can greatly affect the steel's resistance to damage by He irradiation. In this work, we examined the effect of solute solubility on He irradiation damage in nanostructured austenitic steel. i) The chemical contributions are charge transfer and bonding between elements, whereas the mechanical contribution is associated with local distortion due to atomic radius mismatch; ii) The bonds between rare earth elements and bulk atoms are weak and short, resulting in the formation of low-energy potential traps around the rare earth elements, enabling them to trap He. This insight is valuable for understanding the mechanisms of radiation resistance in metals.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155636"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155315","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":"Micromechanical response of SiC-OPyC layers in TRISO fuel particles","authors":"Katherine I. Montoya,&nbsp;Erik G. Herbert,&nbsp;Danny P. Schappel,&nbsp;Christian M. Petrie,&nbsp;Andrew T. Nelson,&nbsp;Tyler J. Gerczak","doi":"10.1016/j.jnucmat.2025.155654","DOIUrl":"10.1016/j.jnucmat.2025.155654","url":null,"abstract":"<div><div>Tristructural isotropic (TRISO)–coated particle fuel is a proposed fuel for multiple advanced reactor concepts. The performance of the particle depends on whether the silicon carbide (SiC) layer remains intact to prevent the release of metallic and gaseous fission products. Mechanical fracture of the SiC layer is a potential failure mode under various fuel configurations and operating environments, including the potential transmission of matrix-originating cracks through TRISO particles. This study uses instrumented indentation techniques on cross-sectioned surrogate particles to examine the mechanical stability of the critical interface between SiC and the outer pyrolytic carbon (OPyC) layer. The observed behavior at the interface is rationalized by examining the radially dependent fracture behavior of the SiC layer and performing a numerical analysis to quantify the residual stresses that develop during the processing and cross-sectioning of the as-fabricated particle. Characterizing the SiC-OPyC interface of surrogate TRISO particles using nanoindentation provides unique insight into the interface's room-temperature residual stress and mechanical stability. The modeling efforts were used to investigate the experimental procedure further, and the results are presented herein to validate this fuel form's potential mechanical failure modes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155654"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155400","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":"Utralow coefficient of thermal expansion in a spheroidized natural flake graphite based isotropic graphite","authors":"Xinlei Cao ,&nbsp;Kun Xu ,&nbsp;Yanan Niu ,&nbsp;Shen Lv ,&nbsp;Ke Shen","doi":"10.1016/j.jnucmat.2024.155544","DOIUrl":"10.1016/j.jnucmat.2024.155544","url":null,"abstract":"<div><div>With the development of high-temperature gas-cooled reactors, the coefficient of thermal expansion (CTE) of nuclear-grade graphite plays an increasingly important role in reactor design. A lower CTE enhances both the integrity of the graphite core structure and reactor efficiency. In this paper, we present a new isotropic graphite grade with a low CTE, utilizing spheroidized natural flake graphite (SFG) as a filler material. The ultralow isotropic CTE of 2.6–2.9 × 10⁻⁶ <em>K</em>⁻¹ in the SFG-based graphite, owing to the ability of the slit-shaped pores within the SFG particles to accommodate cross-plane thermal expansion. To enhance the baking performance of the SFG-based graphite, hybrid fillers of SFG/coke or SFG/microcrystalline graphite (MG) were used to prevent cracking of the green bodies. In particular, the addition of MG prevents cracking without changing the low CTE value of the SFG-based graphite. This research contributes to the development of new graphite materials with low CTE that can be used in nuclear engineering, the semiconductor industry, and other high-temperature environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155544"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170545","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":"Corrosion behavior of 304L stainless steel by wire arc additive manufacturing in liquid lead-bismuth eutectic at 550 °C","authors":"Libo Zhang ,&nbsp;Hao Wang ,&nbsp;Ke Zhao ,&nbsp;Fangqiang Ning ,&nbsp;Tianyi Xu ,&nbsp;Jia Liu ,&nbsp;Hong Yan ,&nbsp;Qiang Zhang","doi":"10.1016/j.jnucmat.2024.155598","DOIUrl":"10.1016/j.jnucmat.2024.155598","url":null,"abstract":"<div><div>The corrosion behaviors of wire arc additive manufacturing 304L stainless steel (WAAM 304L SS) and forged 304L stainless steel (304L SS) were studied in static oxygen-saturated/depleted lead-bismuth eutectic (LBE) at 550 °C for up to 1000 h. It was found that WAAM 304L SS exhibited better resistance to LBE corrosion than 304L SS under both saturated and depleted oxygen conditions due to its optimized chemical composition and microstructure. Three-layer oxide films were formed on both materials in oxygen-saturated LBE, and δ-ferrite showed stronger oxidation resistance than austenite in WAAM 304L SS. The corrosion mechanism of WAAM 304L SS in oxygen-saturated/depleted liquid LBE are discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155598"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170553","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}