Journal of Nuclear Materials最新文献

英文中文

Semi-integral LOCA testing of Cr-coated Optimized ZIRLOTM claddings

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-18 DOI: 10.1016/j.jnucmat.2025.155766

Ioannis Alakiozidis , Marc Lopes Nunes de Sousa , Axel Gauthier , Callum Hunt , Mia Maric , Antoine Ambard , Zaheen Shah , Philipp Frankel

Chromium (Cr)-coatings on zirconium-(Zr) based claddings have emerged as a promising short-term solution to enhance the accident tolerance of fuel assemblies in pressurised water reactors (PWRs) during loss-of-coolant accidents (LOCAs). In this study, we tested a large number (36 rods in total, each 30cm long) of uncoated and Cr-coated Optimized ZIRLO^TM claddings under thermomechanical conditions that closely resemble a real LOCA. A unique experimental apparatus was employed to integrate multiple LOCA effects into a single test sequence, enabling a more accurate prediction of the performance of Cr-coatings and degradation mechanisms of the coated claddings. More specifically, the test sequence included: i) thermal ramping from 350–1200°C under varying internal pressures and heating rates in flowing steam; ii) isothermal steam oxidation at 1200°C for different durations; ii) cooling to 700°C followed by water quenching to 135°C; iv) partial-axial constraint at 135°C with load hold of 540N for 20s. Various characterisation techniques, including optical and scanning electron microscopy (SEM), 3D laser scanning, electron backscattered diffraction (EBSD), hardness testing and hydrogen analysis, were used to characterise the post-LOCA cladding microstructures. We found that Cr-coatings increased the burst temperature of uncoated claddings by ∼ 25–150°C and reduced the strain-to-burst and cladding deformation within 20 mm away from the burst opening. The magnitude of these improvements depended on the initial testing conditions and were more pronounced for the helium-propelled cold spray (HCS) coating, while less pronounced for the nitrogen-propelled CS (NCS) and physical vapour deposition (PVD) coatings. Additionally, we found that Cr-coatings increased the time threshold before significant cladding embrittlement by ∼100–555s compared to uncoated claddings. Finally, we concluded that when multiple LOCA effects are considered, predictions of additional coping time during a LOCA provided by the Cr-coatings are more conservative compared to single-factor tests.

锆（Zr）基覆层上的铬（Cr）涂层已成为一种很有前途的短期解决方案，可提高压水堆（PWR）燃料组件在失冷事故（LOCA）期间的事故耐受性。在这项研究中，我们在与真实 LOCA 非常相似的热机械条件下测试了大量（共 36 根棒，每根 30 厘米长）无涂层和有铬涂层的优化 ZIRLOTM 包壳。采用了一种独特的实验装置，将多种 LOCA 效应整合到一个测试序列中，从而能够更准确地预测铬涂层的性能和涂层包层的降解机制。更具体地说，测试序列包括：i) 在流动蒸汽中以不同的内部压力和加热速率从 350°C 升温到 1200°C；ii) 在 1200°C 进行不同持续时间的等温蒸汽氧化；ii) 冷却到 700°C，然后水淬到 135°C；iv) 在 135°C 进行部分轴向约束，负载保持 540N 20 秒。我们采用了各种表征技术，包括光学和扫描电子显微镜 (SEM)、三维激光扫描、电子背散射衍射 (EBSD)、硬度测试和氢分析，来表征 LOCA 后包层的微观结构。我们发现，Cr 涂层将未涂层包层的爆裂温度提高了 25-150°C 左右，并降低了爆裂应变和距爆裂口 20 毫米范围内的包层变形。这些改善的幅度取决于初始测试条件，氦气推进冷喷（HCS）涂层的改善幅度更大，而氮气推进 CS（NCS）和物理气相沉积（PVD）涂层的改善幅度较小。此外，我们还发现，与未涂层的覆层相比，Cr 涂层将覆层发生明显脆化之前的时间阈值提高了 100-555 秒。最后，我们得出结论，当考虑到多重 LOCA 影响时，与单因素测试相比，Cr 涂层提供的 LOCA 期间额外应对时间的预测更为保守。

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引用次数: 0

Low temperature neutron irradiation stability of Zirconium hydride and Yttrium hydride

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-18 DOI: 10.1016/j.jnucmat.2025.155770

D.J. Sprouster , M. Ouyang , N. Cetiner , P. Negi , A. Sharma , D. Bhardwaj , Y. Huang , X. Hu , K. Shirvan , L.L. Snead

Metal hydrides, including ZrH_x and YH_x, are of particular interest for advanced thermal fission reactors as they have high neutron moderating power and can be used at relatively high temperatures. They have direct applications as core components including as a moderating addition in nuclear fuel, and as neutron reflectors or moderators. Understanding their thermal and irradiation-induced property changes are important to their engineering application. Specifically, evolving metal hydrogen ratios are of critical importance. In this work we discuss the post-irradiation examination of neutron irradiated ZrH_2-x and YH_2-x specimens. We employ multiple characterization techniques including X-ray diffraction, scanning electron microscopy and thermophysical (thermal diffusivity) to determine the irradiation-induced macro- and microstructural evolution as a function of irradiation temperature. We readily quantify degradations in the thermal diffusivity, changes in lattice parameters, and an increase in metallic Zr indicative of hydrogen release in ZrH_2-x specimens. Interestingly, minimal-to-nil change in the metallic Y fraction was quantifiable in the YH_2-x specimens and modest changes in the thermal diffusivity occur for the temperature and dose studied. The loss of hydrogen in the ZrH_2-x samples is related to an apparent irradiation-accelerated desorption of hydrogen by the high ionizing radiation components (gamma, epithermal and fast neutron fluxes) from the in-core neutron irradiation. The most apparent feature from the microstructural analysis for both metal hydrides was a temperature-dependent decrease in the X-ray diffraction peak broadening, attributable to changes in the number and makeup of the two-dimensional defects. These results and trends improve both the fundamental understanding of neutron-solid interactions, and the development of such an important class of core materials.

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引用次数: 0

Continuum-mechanics-based multi-scale modeling of fission gas swelling and release coupling behaviors for UO2 fuels

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-15 DOI: 10.1016/j.jnucmat.2025.155757

Jing Zhang , Feng Yan , Shurong Ding

Under the extreme in-pile environments, gaseous fission products continuously accumulate within nuclear fuels, causing macroscopic fuel swelling and fission gas release. Fission gas swelling and release effects are coupled to each other, related to the diffusion behavior of fission gas atoms within fuel grains. With the evolution of bubbles, the fuels transform into porous structures, degrading their macroscopic thermo-mechanical properties and thereby affecting the overall thermo-mechanical behaviors of the fuel elements. Conducting multi-scale modeling studies on the coupling behaviors of fission gas swelling and release, and achieving accurate predictions of these behaviors, are essential scientific problems and important concerns in reactor engineering design. In this study, the macroscopic volume changes and fission gas release behaviors of porous fuels are considered to be associated with the diffusion of fission gas atom, the growth of inter-granular bubbles, grain recrystallization effects and the creep deformations of the solid skeleton. By considering the contributions of bubble internal pressure, surface tension and external hydrostatic pressure, a multi-scale model describing fission gas swelling and release coupling behaviors is established based on continuum mechanics theory. This model is validated using abundant experimental results. Furthermore, the underlying mechanisms of fission gas swelling and release coupling behaviors are revealed. It is found that the creep deformation of the surrounding skeleton is the primary contributor to fission gas swelling, and creep-related damage of the skeleton appears to be the dominant mechanism for fission gas release and bubble connection. This study can provide technical support for the multi-scale thermo-mechanical behavior analysis of many advanced fuel elements.

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引用次数: 0

Experimental confirmation of first-principles thermal conductivity in Zirconium-doped ThO2

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-15 DOI: 10.1016/j.jnucmat.2025.155756

Ella Kartika Pek , Zilong Hua , Amey Khanolkar , J. Matthew Mann , David B. Turner , Karl Rickert , Timothy A. Prusnick , Marat Khafizov , David H. Hurley , Linu Malakkal

The degradation of thermal conductivity in advanced nuclear fuels due to the accumulation of fission products and irradiation-induced defects is inevitable, and must be considered as part of safety and efficiency analyses of nuclear reactors. This study examines the thermal conductivity of a zirconium-doped ThO₂ crystal, synthesized via the hydrothermal method using a spatial domain thermoreflectance technique. Zirconium is one of the soluble fission products in oxide fuels that can effectively scatter heat-carrying phonons in the crystalline lattice of fuel. Thus, thermal property measurements of zirconium-doped ThO₂ single crystals provide insights into the effects of substitutional zirconium doping, isolated from extrinsic factors such as grain boundary scattering. The experimental results are compared with first-principles calculations of the lattice thermal conductivity of ThO₂, employing an iterative solution of the Peierls-Boltzmann transport equation. Additionally, the non-perturbative Green's function methodology is utilized to compute phonon-point defect scattering rates, accounting for local distortions around point defects, including mass difference changes, interatomic force constants, and structural relaxation. The congruence between the predicted results from first-principles calculations and the measured temperature-dependent thermal conductivity validates the computational methodology. Furthermore, the methodologies employed in this study enable systematic investigations of thermal conductivity reduction by fission products, potentially leading to the development of more accurate fuel performance codes.

{"title":"Experimental confirmation of first-principles thermal conductivity in Zirconium-doped ThO2","authors":"Ella Kartika Pek , Zilong Hua , Amey Khanolkar , J. Matthew Mann , David B. Turner , Karl Rickert , Timothy A. Prusnick , Marat Khafizov , David H. Hurley , Linu Malakkal","doi":"10.1016/j.jnucmat.2025.155756","DOIUrl":"10.1016/j.jnucmat.2025.155756","url":null,"abstract":"<div><div>The degradation of thermal conductivity in advanced nuclear fuels due to the accumulation of fission products and irradiation-induced defects is inevitable, and must be considered as part of safety and efficiency analyses of nuclear reactors. This study examines the thermal conductivity of a zirconium-doped ThO<sub>2</sub> crystal, synthesized via the hydrothermal method using a spatial domain thermoreflectance technique. Zirconium is one of the soluble fission products in oxide fuels that can effectively scatter heat-carrying phonons in the crystalline lattice of fuel. Thus, thermal property measurements of zirconium-doped ThO<sub>2</sub> single crystals provide insights into the effects of substitutional zirconium doping, isolated from extrinsic factors such as grain boundary scattering. The experimental results are compared with first-principles calculations of the lattice thermal conductivity of ThO<sub>2</sub>, employing an iterative solution of the Peierls-Boltzmann transport equation. Additionally, the non-perturbative Green's function methodology is utilized to compute phonon-point defect scattering rates, accounting for local distortions around point defects, including mass difference changes, interatomic force constants, and structural relaxation. The congruence between the predicted results from first-principles calculations and the measured temperature-dependent thermal conductivity validates the computational methodology. Furthermore, the methodologies employed in this study enable systematic investigations of thermal conductivity reduction by fission products, potentially leading to the development of more accurate fuel performance codes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"609 ","pages":"Article 155756"},"PeriodicalIF":2.8,"publicationDate":"2025-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682797","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

Zirconium metal electrodeposition on uranium nitride in molten fluoride salts

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-14 DOI: 10.1016/j.jnucmat.2025.155750

Jarom L. Chamberlain , Hannah K. Patenaude , Amanda L. Musgrove , Rami J. Batrice , Timothy P. Coons , Marisa J. Monreal

Uranium nitride (UN) has application as fuel for nuclear thermal rockets and advanced nuclear reactors. The high melting point, high fissile density, and thermal conductivity of UN makes it an attractive candidate for fuel in these applications. Coating uranium nitride fuel with metal such as zirconium provides additional stability and containment to the UN, promoting its survivability and accident tolerance. This study demonstrates molten salt electrodeposition as a method to deposit a coating of zirconium metal onto a uranium nitride substrate. Cyclic voltammetry was used to characterize the molten salt system and demonstrate the zirconium precursor reduction. Post electrodeposition characterization depicted a zirconium metal coating on the uranium nitride substrate. Preferential growth was observed on one of the substrate interfaces. The average thickness of the coating where preferential growth was depicted was 194 µm.

引用次数: 0

Density functional theory simulation study of Fe solutes in hcp zirconium: Magnetic and electronic properties

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-14 DOI: 10.1016/j.jnucmat.2025.155755

Junting Zhang, Andrew Horsfield, Mark Wenman

Fe is added to improve corrosion resistance of most commercial Zr alloys. This work aims to study Fe solute stability in different interstitial and substitutional sites in hcp α-Zr lattice and Fe solute ferromagnetic properties within these sites using density functional theory (DFT). A relationship between the electronic and magnetic properties of these Fe solutes and their Zr host atom neighbours was found. The stability of the sites, ranked from most to least stable, is as follows: octahedral, substitutional, basal crowdion, basal octahedral, tetrahedral, and basal tetrahedral. An additional off-site substitutional position was examined to evaluate the influence of Fe solute position on the magnetic properties in Zr. The correlation between the stability of interstitial sites and the amount of charge taken from the surrounding Zr atoms was found using Bader charge analysis. From the perspective of magnetic properties, for all tested sites, only the high symmetry Fe substitution remains magnetised in the Zr lattice. Comparison between the local density of states of the Fe defects and their Zr neighbours suggests the interaction between the d-orbitals of Zr and Fe atoms suppresses the local magnetic moment on Fe interstitials.

{"title":"Density functional theory simulation study of Fe solutes in hcp zirconium: Magnetic and electronic properties","authors":"Junting Zhang, Andrew Horsfield, Mark Wenman","doi":"10.1016/j.jnucmat.2025.155755","DOIUrl":"10.1016/j.jnucmat.2025.155755","url":null,"abstract":"<div><div>Fe is added to improve corrosion resistance of most commercial Zr alloys. This work aims to study Fe solute stability in different interstitial and substitutional sites in hcp α-Zr lattice and Fe solute ferromagnetic properties within these sites using density functional theory (DFT). A relationship between the electronic and magnetic properties of these Fe solutes and their Zr host atom neighbours was found. The stability of the sites, ranked from most to least stable, is as follows: octahedral, substitutional, basal crowdion, basal octahedral, tetrahedral, and basal tetrahedral. An additional off-site substitutional position was examined to evaluate the influence of Fe solute position on the magnetic properties in Zr. The correlation between the stability of interstitial sites and the amount of charge taken from the surrounding Zr atoms was found using Bader charge analysis. From the perspective of magnetic properties, for all tested sites, only the high symmetry Fe substitution remains magnetised in the Zr lattice. Comparison between the local density of states of the Fe defects and their Zr neighbours suggests the interaction between the d-orbitals of Zr and Fe atoms suppresses the local magnetic moment on Fe interstitials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"609 ","pages":"Article 155755"},"PeriodicalIF":2.8,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682796","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

Mechanical and microstructural characterization of thermally grown alumina on Kanthal APMT

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-13 DOI: 10.1016/j.jnucmat.2025.155753

Peter Beck , Md. Mehadi Hassan , Arjen van Veelen , Tarik Saleh , Christopher Matthews , Erofili Kardoulaki , Benjamin Eftink

Thermally grown α-alumina layers on Kanthal APMT were characterized and mechanically tested. Different growth durations and post growth heat treatments were evaluated. Gaugeless ring pull tests were conducted with digital image correlation to measure the spallation strain of the oxide layer, providing tens of separate measurements per test. The strain at oxide spallation failure had significant spread with average failure occurring at 0.5 - 2 % strain in both tension and compression. It was also found that the failure strain was dependent on oxide thickness with the thicker 2 and 3 µm oxides tending to fail at lower strains. The thicker oxide layers also experienced more catastrophic failure, with larger spallation flakes leaving more of the base metal exposed. Transmission Kikuchi diffraction was used to characterize the microstructure of the oxide layer. All oxide layers showed columnar grains with grain boundaries extending across the entire thickness of the oxide.

引用次数: 0

Crystallographic and Thermodynamic Insights into Preferential Hydride Precipitation Sites in Zr-2.5Nb Alloy

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-13 DOI: 10.1016/j.jnucmat.2025.155752

Haoyu Zhai , Xiaoqing Shang , Minglang Li , Hao Lin , Ling Li , Yibin Tang , Shengyi Zhong

The Zr-2.5Nb alloy, widely used in nuclear applications, exhibits significant susceptibility to hydrogen-induced embrittlement due to hydride precipitation. This study investigates the preferential sites and mechanisms of hydride precipitation in Zr-2.5Nb alloys using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Results show a predominance of intergranular hydrides, with grain boundaries (GBs) serving as the primary nucleation sites. Misorientation and GB energy exhibited a weak influence on intergranular hydride precipitation, while the interaction angle between basal planes and GBs (

α_{G B - B P}

) was found to determine hydride precipitation behavior. A modified thermodynamic model was developed to elucidate the interplay between GB energy,

α_{G B - B P}

, and hydride precipitation. Additionally, the lamellar β-Zr phase at GBs promotes hydride formation, which largely explains the weak correlation observed between misorientation and intergranular hydride precipitation. These findings provide insights into mitigating hydride-induced degradation in Zr alloys for enhanced performance in nuclear environments.

{"title":"Crystallographic and Thermodynamic Insights into Preferential Hydride Precipitation Sites in Zr-2.5Nb Alloy","authors":"Haoyu Zhai , Xiaoqing Shang , Minglang Li , Hao Lin , Ling Li , Yibin Tang , Shengyi Zhong","doi":"10.1016/j.jnucmat.2025.155752","DOIUrl":"10.1016/j.jnucmat.2025.155752","url":null,"abstract":"<div><div>The Zr-2.5Nb alloy, widely used in nuclear applications, exhibits significant susceptibility to hydrogen-induced embrittlement due to hydride precipitation. This study investigates the preferential sites and mechanisms of hydride precipitation in Zr-2.5Nb alloys using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Results show a predominance of intergranular hydrides, with grain boundaries (GBs) serving as the primary nucleation sites. Misorientation and GB energy exhibited a weak influence on intergranular hydride precipitation, while the interaction angle between basal planes and GBs (<span><math><msub><mi>α</mi><mrow><mi>G</mi><mi>B</mi><mo>−</mo><mi>B</mi><mi>P</mi></mrow></msub></math></span>) was found to determine hydride precipitation behavior. A modified thermodynamic model was developed to elucidate the interplay between GB energy, <span><math><msub><mi>α</mi><mrow><mi>G</mi><mi>B</mi><mo>−</mo><mi>B</mi><mi>P</mi></mrow></msub></math></span>, and hydride precipitation. Additionally, the lamellar β-Zr phase at GBs promotes hydride formation, which largely explains the weak correlation observed between misorientation and intergranular hydride precipitation. These findings provide insights into mitigating hydride-induced degradation in Zr alloys for enhanced performance in nuclear environments.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"609 ","pages":"Article 155752"},"PeriodicalIF":2.8,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682792","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

Study on the structure of Li2TiO3-Li4SiO4 tritium breeder and its performance under thermal cycle loading with compression pressure

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-13 DOI: 10.1016/j.jnucmat.2025.155726

Anjie Yang, Yifu Xu, Kaixuan Zhu, Qilai Zhou

Li₂TiO₃-xLi₄SiO₄ (x = 0.5, 1, 2) pebbles were fabricated to investigate the mechanisms of microstructure variation for the ceramic with different phase ratios. The structure stability of the pebbles under simulated working conditions in fusion reactors was examined. The crush load of Li₂TiO₃-xLi₄SiO₄ (x = 2) pebbles reached 129.6 N. The ceramic with this phase ratio has a higher activation energy (E_a) for grain growth at high temperatures, which suppresses excessive grain growth. The synergetic effects of high temperature, thermal cycle loading, and compression pressure on the structure stability of pebbles were investigated. There was no significant change in the structure and mechanical properties of the pebbles after heating at a constant temperature under compression pressure. However, the strength of the pebbles deteriorated rapidly when exposed to thermal cycle loading with compression pressure. These results suggested that the simultaneous exposure to compression pressure and thermal cycle loading would accelerate the deterioration of the ceramic structure, which should attract more attention from the viewpoint of applying the pebbles in fusion reactors.

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引用次数: 0

Deep potential molecular dynamics simulation of local structure and properties of LiCl-KCl-CsCl-LaCl3 molten salt

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials

Pub Date : 2025-03-12 DOI: 10.1016/j.jnucmat.2025.155749

Changzu Zhu, Jia Song, Yujiao Wang, Haofeng Luo, Yuncong Ding, Wentao Zhou, Yafei Wang

LiCl-KCl-CsCl molten salt is regarded as an ideal electrolyte for the pyroprocessing of spent nuclear fuel due to the lower melting point compared to molten salts studied in the mainstream. In this work, the local structure and properties of LiCl-KCl-CsCl-LaCl₃ molten salts were systematically investigated over the temperature range of 573–813 K using deep potential molecular dynamics simulations. The short-range and intermediate-range ordering, along with the coordination environment of La³⁺ and their dependence on temperature and LaCl₃ concentration, were analyzed based on radial distribution functions and structure factors. La³⁺ predominantly exists as 6-coordinated clusters in the melt because of its low free energy. As temperature and LaCl₃ concentration rise, the short-range ordering of the melt decreases due to the weakened interactions between cations and Cl^-, whereas the intermediate-range ordering exhibits an increasing trend. The variation in intermediate-range ordering is determined by both the Cl^--decorated La³⁺ networks and the La-La networks. Moreover, a series of properties of LiCl-KCl-CsCl-LaCl₃ melts were evaluated, including the self-diffusion coefficient, viscosity, ionic conductivity, heat capacity, thermal expansion coefficient, and thermal conductivity. With the continuous La enrichment in the salt, LiCl-KCl-CsCl molten salt demonstrates excellent electrical conductivity and thermophysical properties, highlighting its advantages and potential as a superior alternative for LiCl-KCl molten salt in pyroprocessing.

{"title":"Deep potential molecular dynamics simulation of local structure and properties of LiCl-KCl-CsCl-LaCl3 molten salt","authors":"Changzu Zhu, Jia Song, Yujiao Wang, Haofeng Luo, Yuncong Ding, Wentao Zhou, Yafei Wang","doi":"10.1016/j.jnucmat.2025.155749","DOIUrl":"10.1016/j.jnucmat.2025.155749","url":null,"abstract":"<div><div>LiCl-KCl-CsCl molten salt is regarded as an ideal electrolyte for the pyroprocessing of spent nuclear fuel due to the lower melting point compared to molten salts studied in the mainstream. In this work, the local structure and properties of LiCl-KCl-CsCl-LaCl<sub>3</sub> molten salts were systematically investigated over the temperature range of 573–813 K using deep potential molecular dynamics simulations. The short-range and intermediate-range ordering, along with the coordination environment of La<sup>3+</sup> and their dependence on temperature and LaCl<sub>3</sub> concentration, were analyzed based on radial distribution functions and structure factors. La<sup>3+</sup> predominantly exists as 6-coordinated clusters in the melt because of its low free energy. As temperature and LaCl<sub>3</sub> concentration rise, the short-range ordering of the melt decreases due to the weakened interactions between cations and Cl<sup>-</sup>, whereas the intermediate-range ordering exhibits an increasing trend. The variation in intermediate-range ordering is determined by both the Cl<sup>-</sup>-decorated La<sup>3+</sup> networks and the La-La networks. Moreover, a series of properties of LiCl-KCl-CsCl-LaCl<sub>3</sub> melts were evaluated, including the self-diffusion coefficient, viscosity, ionic conductivity, heat capacity, thermal expansion coefficient, and thermal conductivity. With the continuous La enrichment in the salt, LiCl-KCl-CsCl molten salt demonstrates excellent electrical conductivity and thermophysical properties, highlighting its advantages and potential as a superior alternative for LiCl-KCl molten salt in pyroprocessing.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"609 ","pages":"Article 155749"},"PeriodicalIF":2.8,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143643137","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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