Journal of Nuclear Materials最新文献

{"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}

{"title":"Corrigendum to “Non-electric mass transfer between stainless steel 316H and glassy carbon in NaF-KF-UF4 salt” [Journal of Nuclear Materials 604 (2025) 155534]","authors":"Jaewoo Park,&nbsp;Jinsuo Zhang","doi":"10.1016/j.jnucmat.2024.155557","DOIUrl":"10.1016/j.jnucmat.2024.155557","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155557"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170987","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":"Inversion of the fracture toughness of zirconium alloy cladding interface in nuclear fuel using splitting method via general regression neural network","authors":"Yubo Zhou ,&nbsp;Yingxuan Dong ,&nbsp;Haojun Ma ,&nbsp;Junnan Lv ,&nbsp;Qun Li","doi":"10.1016/j.jnucmat.2024.155573","DOIUrl":"10.1016/j.jnucmat.2024.155573","url":null,"abstract":"<div><div>For nuclear fuel elements, the interface mechanical properties of zirconium alloy cladding is critical to the safety and reliability of reactors. However, due to the small thickness of the fuel plates (&lt;2 mm), accurately capturing the behavior of interface cracks is challenging, complicates the measurement of interface fracture toughness. This study developed a data-driven inversion method using the generalized regression neural network (GRNN) to rapidly and accurately determine the fracture toughness of zirconium alloy cladding interface. The database was established by combining splitting experiments with numerical simulations. The cohesive zone model was utilized to accurately simulate crack propagation paths and fracture modes in numerical simulations. The influences of key parameters such as cohesive strength, stiffness, and interface fracture energy were analyzed in detail. After extensive training, the prediction model accurately forecasted the interface fracture toughness. The results indicate that the proposed GRNN-based inversion approach is feasible and effective for predicting the fracture toughness of zirconium alloy cladding interface, and can be extended to determinations of other mechanical properties in the nuclear fuel element.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155573"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171593","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":"Thermochemical bounds on UCO TRISO kernel compositions","authors":"M. Poschmann,&nbsp;D. Wojtaszek,&nbsp;E. Geiger","doi":"10.1016/j.jnucmat.2024.155537","DOIUrl":"10.1016/j.jnucmat.2024.155537","url":null,"abstract":"<div><div>Highly-detailed equilibrium thermodynamics calculations were used to derive bounds on safe fresh UCO TRISO fuel kernel chemistries, in terms of the fraction of UC_<em>x</em>. The lower bound on UC_<em>x</em> was chosen to limit CO pressure, and the upper bound to ensure oxidation of radionuclide fission and transmutation products of concern. The bounds can be calculated with respect to both target burnup and fuel temperature. Calculated bounds for a demonstration fuel design similar to the AGR-2 experiment are in good agreement with previously-calculated lower bounds and experimental evidence used to determine upper bounds, and indicate that reference fuel designs such as AGR-2 and MHTGR-350 are well-within the safe composition region even for burnup as high as 21.3% FIMA. Predicted speciations of cesium, palladium, silver, barium, strontium, and zirconium were examined, and recommendations for thermodynamic assessments of relevant systems provided.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155537"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171598","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":"Dispersed barrier hardening modeling on depth-distributed helium bubbles in iron-based alloys","authors":"Xinrun Chen ,&nbsp;Tatsuya Suzuki ,&nbsp;Phongsakorn Prak Tom ,&nbsp;Bo Li ,&nbsp;Zongda Yang ,&nbsp;Sho Kano ,&nbsp;Takuya Yamamoto ,&nbsp;Kenta Murakami","doi":"10.1016/j.jnucmat.2025.155608","DOIUrl":"10.1016/j.jnucmat.2025.155608","url":null,"abstract":"<div><div>The present study investigates the irradiation hardening in pure Fe and Fe-0.3 wt.% Si alloys after He⁺ ion implantation tests in order to obtain a representative value for the barrier strength (α-value) of helium bubbles in model ferritic alloys. 160 keV He⁺ ion implantation was performed at 400 ± 2 °C with a fluence up to 3.6 × 10¹⁶ cm^-2 at a flux of 4.0 × 10¹³ cm^-2∙s^-1. Hardness was measured using the nanoindentation technique at different indentation depths, and the depth-distributed helium bubbles were observed by transmission electron microscopy (TEM). The results of the hardness measurement demonstrated a significant damage gradient effect in both He⁺ implanted specimens. This could be attributed to the presence of depth-distributed helium bubbles. Additionally, a significant dose dependence on swelling was observed in both alloys, suggesting that the primary parameter governing the evolution of helium bubbles is the dose. While the traditional dispersed barrier hardening (DBH) model could be employed to evaluate the α-value based on helium bubble-induced hardening, the distribution of dispersed barriers should be uniform. Consequently, we proposed a more efficacious application of the dispersed barrier hardening model to describe the depth-distributed helium bubble-induced hardening in iron-based materials by integrating the nanoindentation technique with TEM. The α-value for helium bubbles of approximately 2∼3 nm in diameter in model ferritic alloys is estimated to be approximately 0.15 to 0.16 based on the revised DBH model, which is consistent with the value observed in annealed F82H of neutron irradiation. Furthermore, this study suggested that the much greater hardness increase and swelling rate of implanted Fe-0.3Si than that of pure iron is due to the additive silicon, which has a strong ability to inhibit vacancy diffusion in the evolution of helium bubbles.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155608"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154878","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}