Journal of Nuclear Materials最新文献

英文中文

Damage behavior of He-irradiated sintered SiC at high temperatures

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

Journal of Nuclear Materials

Pub Date : 2025-04-21 DOI: 10.1016/j.jnucmat.2025.155839

Jintao Zhang , Zhongzheng Wu , Jiale Huang , Jun Li , Haiyuan Wei , Tongmin Zhang , Tomas Polcar , Nabil Daghbouj , Bingsheng Li

Understanding the effects of high-temperature helium (He) irradiation on the damage behavior of sintered silicon carbide (SiC) is crucial for assessing the material's stability in advanced nuclear reactors. In this study, we investigate the impact of 230 keV He ions on SiC at temperatures of 800 °C and 1000 °C, utilizing three different irradiation fluences: 2 × 10¹⁶/cm², 4 × 10¹⁶/cm², and 1.6 × 10¹⁷/cm². Raman spectroscopy and transmission electron microscopy were employed to analyze various damage features, including irradiation-induced lattice strain, platelet formation, dislocation loops, and helium bubbles. Our findings indicate that over-pressurized platelets predominantly formed on the (0001) plane, with a limited number of dislocation loops detected nearby. In contrast, numerous black spot defects were observed near grain boundaries, where platelets were absent. This variation in defect distribution underscores the unique damage behavior associated with high-temperature He irradiation. The insights gained from this study are essential for understanding the structural changes and integrity of SiC materials under conditions relevant to nuclear reactor applications.

{"title":"Damage behavior of He-irradiated sintered SiC at high temperatures","authors":"Jintao Zhang , Zhongzheng Wu , Jiale Huang , Jun Li , Haiyuan Wei , Tongmin Zhang , Tomas Polcar , Nabil Daghbouj , Bingsheng Li","doi":"10.1016/j.jnucmat.2025.155839","DOIUrl":"10.1016/j.jnucmat.2025.155839","url":null,"abstract":"<div><div>Understanding the effects of high-temperature helium (He) irradiation on the damage behavior of sintered silicon carbide (SiC) is crucial for assessing the material's stability in advanced nuclear reactors. In this study, we investigate the impact of 230 keV He ions on SiC at temperatures of 800 °C and 1000 °C, utilizing three different irradiation fluences: 2 × 10<sup>16</sup>/cm<sup>2</sup>, 4 × 10<sup>16</sup>/cm<sup>2</sup>, and 1.6 × 10<sup>17</sup>/cm<sup>2</sup>. Raman spectroscopy and transmission electron microscopy were employed to analyze various damage features, including irradiation-induced lattice strain, platelet formation, dislocation loops, and helium bubbles. Our findings indicate that over-pressurized platelets predominantly formed on the (0001) plane, with a limited number of dislocation loops detected nearby. In contrast, numerous black spot defects were observed near grain boundaries, where platelets were absent. This variation in defect distribution underscores the unique damage behavior associated with high-temperature He irradiation. The insights gained from this study are essential for understanding the structural changes and integrity of SiC materials under conditions relevant to nuclear reactor applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155839"},"PeriodicalIF":2.8,"publicationDate":"2025-04-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860188","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

Radiation response of inconel-Cu multimetallic layered composites: Role of alloy chemistry

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

Journal of Nuclear Materials

Pub Date : 2025-04-19 DOI: 10.1016/j.jnucmat.2025.155837

Rajesh Ramesh , Daniel Schwen , Sara Neshani , Keivan Davami , Kasra Momeni

The reliability and longevity of fusion reactors' current drive systems are essential for sustained operation under high neutron fluence, requiring materials that maintain high strength and conductivity while resisting high irradiation doses. Here, we investigate the stability and structural integrity of under irradiation using Molecular Dynamics simulations. The mechanochemistry of the constituent elements and their role in the radiation resistance is investigated by considering five variants of Inconel, i.e., Incoloy 800H (Ni₃₂Cr₂₁Fe₄₇), Inconel 625 (Ni₇₂Cr₂₃Fe₅), Inconel 690 (Ni₅₈Cr₃₁Fe₁₁), Inconel 718 (Ni₅₅Cr₂₁Fe₂₄), and Inconel X-750 (Ni₇₇Cr₁₄Fe₉). We revealed three linear, exponential, and plateau stages in the relationship between Frenkel Pair (FP) defect density and radiation damage. Furthermore, our results indicate a high Fe concentration reduces diffusivity between the two metallic layers, while a high concentration of Cr, with its low migration energy barrier, increases diffusivity. Among considered composites, the Incoloy 800H (Ni₃₂Cr₂₁Fe₄₇) shows the highest radiation resistance. FP defect clustering planes are revealed in both Inconel and Cu, while the formation of Stacking Faults (SF) and Lomer-Cottrell (LC) locks are also observed on the Inconel side; we revealed that the shear stress determines the orientation of the FP defect clustering planes.

{"title":"Radiation response of inconel-Cu multimetallic layered composites: Role of alloy chemistry","authors":"Rajesh Ramesh , Daniel Schwen , Sara Neshani , Keivan Davami , Kasra Momeni","doi":"10.1016/j.jnucmat.2025.155837","DOIUrl":"10.1016/j.jnucmat.2025.155837","url":null,"abstract":"<div><div>The reliability and longevity of fusion reactors' current drive systems are essential for sustained operation under high neutron fluence, requiring materials that maintain high strength and conductivity while resisting high irradiation doses. Here, we investigate the stability and structural integrity of under irradiation using Molecular Dynamics simulations. The mechanochemistry of the constituent elements and their role in the radiation resistance is investigated by considering five variants of Inconel, i.e., Incoloy 800H (Ni<sub>32</sub>Cr<sub>21</sub>Fe<sub>47</sub>), Inconel 625 (Ni<sub>72</sub>Cr<sub>23</sub>Fe<sub>5</sub>), Inconel 690 (Ni<sub>58</sub>Cr<sub>31</sub>Fe<sub>11</sub>), Inconel 718 (Ni<sub>55</sub>Cr<sub>21</sub>Fe<sub>24</sub>), and Inconel X-750 (Ni<sub>77</sub>Cr<sub>14</sub>Fe<sub>9</sub>). We revealed three linear, exponential, and plateau stages in the relationship between Frenkel Pair (FP) defect density and radiation damage. Furthermore, our results indicate a high Fe concentration reduces diffusivity between the two metallic layers, while a high concentration of Cr, with its low migration energy barrier, increases diffusivity. Among considered composites, the Incoloy 800H (Ni<sub>32</sub>Cr<sub>21</sub>Fe<sub>47</sub>) shows the highest radiation resistance. FP defect clustering planes are revealed in both Inconel and Cu, while the formation of Stacking Faults (SF) and Lomer-Cottrell (LC) locks are also observed on the Inconel side; we revealed that the shear stress determines the orientation of the FP defect clustering planes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155837"},"PeriodicalIF":2.8,"publicationDate":"2025-04-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860187","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

Micro-scale signature for suppressing fragile fracture in high entropy alloys under irradiation

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

Journal of Nuclear Materials

Pub Date : 2025-04-17 DOI: 10.1016/j.jnucmat.2025.155834

Peng-wei Wang , Babafemi Malomo , Shu-quan Chang , Liang Yang

Compared with traditional alloys, high entropy alloys (HEAs) have better resistance to irradiation embrittlement and hardening, which continue to gain significant attention as promising high-end structural materials, but up until now, the underpinnings of suppressing brittle failure are yet to be revealed, limiting their application. Hence, this study proposes a molecular dynamics framework that can apprehend the evolutions of nano-scale dislocations and micron-sized shear bands from a microstructural evolution-energetics standpoint to elucidate deformation mechanisms in FeNiCrCuAl HEAs under irradiation. Accordingly, prototypic models (0 dpa, 0.02 dpa and 0.2 dpa) of the HEA, indicated an ultimate tensile strength at equivalent strain point of 4.7 % but as strengths declined with the progression of strain, multiple irradiations provoked intense atomic-dislocation interactions by which higher dislocation densities stimulated high-energy dislocation intersects for an amplified work-hardening effect. The evolutions of dislocation density with variations in average atomic energies precipitated distinctive shear band mechanisms characterized by multiple shear bands propagations along 45° and 135° directions in all of the models, and by the atomic level internal stresses constraint on atomic mobility under negative pressure, lowered atomic energies induced by intensified irradiations evolved a phenomenal cross-blocking effect of fewer multiple propagating shear bands to indicate higher ultimate tensile strength and enhanced resistance to fragile failure in the HEAs. Thus, by capturing the dislocation-shear band mechanism under irradiated energy potential landscape, the correlation between micro-scale structural evolutions and mechanical behavior was established in unraveling the fragile failure phenomenon in HEAs.

{"title":"Micro-scale signature for suppressing fragile fracture in high entropy alloys under irradiation","authors":"Peng-wei Wang , Babafemi Malomo , Shu-quan Chang , Liang Yang","doi":"10.1016/j.jnucmat.2025.155834","DOIUrl":"10.1016/j.jnucmat.2025.155834","url":null,"abstract":"<div><div>Compared with traditional alloys, high entropy alloys (HEAs) have better resistance to irradiation embrittlement and hardening, which continue to gain significant attention as promising high-end structural materials, but up until now, the underpinnings of suppressing brittle failure are yet to be revealed, limiting their application. Hence, this study proposes a molecular dynamics framework that can apprehend the evolutions of nano-scale dislocations and micron-sized shear bands from a microstructural evolution-energetics standpoint to elucidate deformation mechanisms in FeNiCrCuAl HEAs under irradiation. Accordingly, prototypic models (0 dpa, 0.02 dpa and 0.2 dpa) of the HEA, indicated an ultimate tensile strength at equivalent strain point of 4.7 % but as strengths declined with the progression of strain, multiple irradiations provoked intense atomic-dislocation interactions by which higher dislocation densities stimulated high-energy dislocation intersects for an amplified work-hardening effect. The evolutions of dislocation density with variations in average atomic energies precipitated distinctive shear band mechanisms characterized by multiple shear bands propagations along 45° and 135° directions in all of the models, and by the atomic level internal stresses constraint on atomic mobility under negative pressure, lowered atomic energies induced by intensified irradiations evolved a phenomenal cross-blocking effect of fewer multiple propagating shear bands to indicate higher ultimate tensile strength and enhanced resistance to fragile failure in the HEAs. Thus, by capturing the dislocation-shear band mechanism under irradiated energy potential landscape, the correlation between micro-scale structural evolutions and mechanical behavior was established in unraveling the fragile failure phenomenon in HEAs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155834"},"PeriodicalIF":2.8,"publicationDate":"2025-04-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143854815","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

Development of boron-based thermal neutron absorbers for short-pulsed MW-class neutron sources – how can the pre-decoupling function be enhanced?

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

Journal of Nuclear Materials

Pub Date : 2025-04-16 DOI: 10.1016/j.jnucmat.2025.155779

Toshifumi Okutomi , Makoto Teshigawara , Sota Sato , Stephen Gallimore , Robert Bewley , Motoki Ooi , Masahide Harada , Shigeru Kuramoto

Decouplers (thermal neutron absorbers) are used for pulse shaping of neutron beams in pulsed neutron sources, contributing to higher resolution of neutron instruments. We are developing a boron (B)-based decoupler material for MW pulsed neutron sources, focusing on the pre-decoupling function to suppress material embrittlement due to the (n, α) reaction of B by adding other higher thermal-neutron absorption material (gadolinium (Gd)). The challenge is to develop a material in which B and Gd are uniformly dispersed. In the development of sintered materials focusing on the hot isostatic pressing (HIP) method, the possibility of further enhancing the pre-decoupling function was obtained under HIP temperature conditions from above 893 K to below the melting point.

去耦器（热中子吸收器）用于脉冲中子源中中子束的脉冲整形，有助于提高中子仪器的分辨率。我们正在为 MW 脉冲中子源开发一种基于硼（B）的去耦合器材料，重点是通过添加其他更高的热中子吸收材料（钆（Gd））来实现预去耦合功能，以抑制因 B 的（n，α）反应而导致的材料脆化。目前的挑战是开发一种 B 和 Gd 均匀分散的材料。在以热等静压（HIP）方法为重点的烧结材料开发过程中，在 HIP 温度从 893 K 以上到熔点以下的条件下，获得了进一步增强预解耦功能的可能性。

引用次数: 0

Corrosion behavior of Ti-Grade2 dissolver material in nitric acid containing fluoride ions

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

Journal of Nuclear Materials

Pub Date : 2025-04-13 DOI: 10.1016/j.jnucmat.2025.155833

R. Priya , K. Kaliraj , S. Ningshen

The Ti-grade2 is used as the candidate dissolver material for reprocessing the nuclear-spent fuels in the aqueous-based Plutonium Uranium Recovery by EXtraction (PUREX) reprocessing method. Fluoride is used to enhance the dissolution of high-burn-up fuel. However, fluoride is known to severely degrade the corrosion behaviour of commercially pure Ti (CP-Ti) in reprocessing nitric acid. The present work focuses on understanding the corrosion behaviour of Ti-grade2 dissolver material in nitric acid medium containing fluoride ions with and without complexing agent Al(NO₃)₃ at 11.5 M and 1 M HNO₃ by electrochemical and boiling nitric acid studies. The potentiodynamic polarization results revealed inferior corrosion resistance with active corrosion potential and higher passive current density in 11.5 M HNO₃ + 0.05 M NaF compared to 11.5 M HNO₃. Moreover, the deterioration of corrosion resistance was more pronounced with increasing temperature and fluoride concentration in 11.5 M HNO₃. Spontaneous passivation behavior was observed under all conditions tested in this acid concentration. In contrast, the presence of fluoride in 1 M HNO₃ induced an active-passive transition, characterized by a negative shift in corrosion potential and an increase in passive current density. Corrosion mitigation of titanium was found to be effective in a nitric acid medium containing fluoride and a complexing agent. A significantly higher corrosion rate was observed in 1 M HNO₃ + 0.05 M NaF compared to 11.5 M HNO₃ + 0.05 M NaF during the boiling nitric acid test. SEM, XPS, and AFM analyses supported these findings. The corrosion mechanism by which fluoride influences the corrosion resistance of CP-Ti in both 1 M and 11.5 M nitric acid concentrations was proposed. Additionally, this study provides valuable insights for the use of CP-Ti as a dissolver material in a nuclear reprocessing environment involving nitric acid and fluoride.

{"title":"Corrosion behavior of Ti-Grade2 dissolver material in nitric acid containing fluoride ions","authors":"R. Priya , K. Kaliraj , S. Ningshen","doi":"10.1016/j.jnucmat.2025.155833","DOIUrl":"10.1016/j.jnucmat.2025.155833","url":null,"abstract":"<div><div>The Ti-grade2 is used as the candidate dissolver material for reprocessing the nuclear-spent fuels in the aqueous-based Plutonium Uranium Recovery by EXtraction (PUREX) reprocessing method. Fluoride is used to enhance the dissolution of high-burn-up fuel. However, fluoride is known to severely degrade the corrosion behaviour of commercially pure Ti (CP-Ti) in reprocessing nitric acid. The present work focuses on understanding the corrosion behaviour of Ti-grade2 dissolver material in nitric acid medium containing fluoride ions with and without complexing agent Al(NO<sub>3</sub>)<sub>3</sub> at 11.5 M and 1 M HNO<sub>3</sub> by electrochemical and boiling nitric acid studies. The potentiodynamic polarization results revealed inferior corrosion resistance with active corrosion potential and higher passive current density in 11.5 M HNO<sub>3</sub> + 0.05 M NaF compared to 11.5 M HNO<sub>3</sub>. Moreover, the deterioration of corrosion resistance was more pronounced with increasing temperature and fluoride concentration in 11.5 M HNO<sub>3</sub>. Spontaneous passivation behavior was observed under all conditions tested in this acid concentration. In contrast, the presence of fluoride in 1 M HNO<sub>3</sub> induced an active-passive transition, characterized by a negative shift in corrosion potential and an increase in passive current density. Corrosion mitigation of titanium was found to be effective in a nitric acid medium containing fluoride and a complexing agent. A significantly higher corrosion rate was observed in 1 M HNO<sub>3</sub> + 0.05 M NaF compared to 11.5 M HNO<sub>3</sub> + 0.05 M NaF during the boiling nitric acid test. SEM, XPS, and AFM analyses supported these findings. The corrosion mechanism by which fluoride influences the corrosion resistance of CP-Ti in both 1 M and 11.5 M nitric acid concentrations was proposed. Additionally, this study provides valuable insights for the use of CP-Ti as a dissolver material in a nuclear reprocessing environment involving nitric acid and fluoride.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155833"},"PeriodicalIF":2.8,"publicationDate":"2025-04-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143837835","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

Assessing the viability of a new subsize tensile specimen geometry for evaluation of structural nuclear and additively manufactured materials 评估用于评估核结构材料和快速成型材料的新型亚尺寸拉伸试样几何形状的可行性

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

Journal of Nuclear Materials

Pub Date : 2025-04-12 DOI: 10.1016/j.jnucmat.2025.155831

David Collins , Maxim Gussev , Stephen Taller , T.S. Byun , Caleb Massey

The use of subsize specimens in nuclear materials testing has been a subject of ongoing interest due to radiation safety concerns and resource conservation needs. It is also of interest in additive manufacturing (AM), also known as 3D-printing, as subsize specimens can more accurately represent the behavior of the small, intricate geometries often produced using AM. A novel, extremely small geometry, called the Subsize Teeny (SST) was recently developed and is of interest for implementation. Given its extraordinarily small size and the complexities associated with subsize specimen testing, adequate vetting of this geometry is necessary to ensure data quality. A variety of unirradiated nuclear structural materials were tested in the SST geometry and compared against the well-established SSJ3 geometry. In addition, two case studies implementing the SST as a screening geometry for AM materials were also conducted. The question of SST viability was found to be highly nuanced and will often be dependent on the context or application in question. It was determined, however, that the SST is a largely invalid geometry for exceptionally coarse-grained materials or in cases where the physical defect volume equals or exceeds 0.1 % or where the specimen machining parameters result in significant surface alterations. On the other hand, it was determined that the SST may be employed with confidence if the test material is nearly or totally free of physical defects, isotropic, demonstrates homogeneous plastic deformation, and possesses a fine-grained, nearly or totally homogeneous microstructure with at least twelve slip systems.

{"title":"Assessing the viability of a new subsize tensile specimen geometry for evaluation of structural nuclear and additively manufactured materials","authors":"David Collins , Maxim Gussev , Stephen Taller , T.S. Byun , Caleb Massey","doi":"10.1016/j.jnucmat.2025.155831","DOIUrl":"10.1016/j.jnucmat.2025.155831","url":null,"abstract":"<div><div>The use of subsize specimens in nuclear materials testing has been a subject of ongoing interest due to radiation safety concerns and resource conservation needs. It is also of interest in additive manufacturing (AM), also known as 3D-printing, as subsize specimens can more accurately represent the behavior of the small, intricate geometries often produced using AM. A novel, extremely small geometry, called the Subsize Teeny (SST) was recently developed and is of interest for implementation. Given its extraordinarily small size and the complexities associated with subsize specimen testing, adequate vetting of this geometry is necessary to ensure data quality. A variety of unirradiated nuclear structural materials were tested in the SST geometry and compared against the well-established SSJ3 geometry. In addition, two case studies implementing the SST as a screening geometry for AM materials were also conducted. The question of SST viability was found to be highly nuanced and will often be dependent on the context or application in question. It was determined, however, that the SST is a largely invalid geometry for exceptionally coarse-grained materials or in cases where the physical defect volume equals or exceeds 0.1 % or where the specimen machining parameters result in significant surface alterations. On the other hand, it was determined that the SST may be employed with confidence if the test material is nearly or totally free of physical defects, isotropic, demonstrates homogeneous plastic deformation, and possesses a fine-grained, nearly or totally homogeneous microstructure with at least twelve slip systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155831"},"PeriodicalIF":2.8,"publicationDate":"2025-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143843402","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

Polarizable force fields for the structural and thermophysical properties of molten actinide chlorides

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

Journal of Nuclear Materials

Pub Date : 2025-04-11 DOI: 10.1016/j.jnucmat.2025.155822

Giovanni Pireddu , Agustin Salcedo , Hugo Sauzet , Sylvie Delpech , David Lambertin , Timothée Kooyman

New nuclear technologies could involve the extensive use of molten salts, including actinide halides. Despite their importance, several practical challenges limit experimental measurements, resulting in knowledge gaps for structural and thermophysical properties. In this work, new polarizable force fields based on ab initio calculations for the simulation of molten actinide chlorides are introduced. The new force fields are used to compute structural properties, density, heat capacity, and isothermal compressibility of pure actinide molten salts (ThCl₄, PaCl₃, NpCl₃, AmCl₃, CmCl₃) at various temperatures. UCl₃ and PuCl₃, which were parameterized in previous works, are also included. The results are discussed in the context of already existing theoretical and experimental datasets, showing good agreement with the literature. Predictions are extended to systems not considered in previous works. Notably, the results highlight the peculiarity of ThCl₄ compared to actinide trichlorides in terms of structural and thermophysical properties. The new force fields can be used in future works for the simulation of molten salts mixtures containing actinides.

{"title":"Polarizable force fields for the structural and thermophysical properties of molten actinide chlorides","authors":"Giovanni Pireddu , Agustin Salcedo , Hugo Sauzet , Sylvie Delpech , David Lambertin , Timothée Kooyman","doi":"10.1016/j.jnucmat.2025.155822","DOIUrl":"10.1016/j.jnucmat.2025.155822","url":null,"abstract":"<div><div>New nuclear technologies could involve the extensive use of molten salts, including actinide halides. Despite their importance, several practical challenges limit experimental measurements, resulting in knowledge gaps for structural and thermophysical properties. In this work, new polarizable force fields based on ab initio calculations for the simulation of molten actinide chlorides are introduced. The new force fields are used to compute structural properties, density, heat capacity, and isothermal compressibility of pure actinide molten salts (ThCl<sub>4</sub>, PaCl<sub>3</sub>, NpCl<sub>3</sub>, AmCl<sub>3</sub>, CmCl<sub>3</sub>) at various temperatures. UCl<sub>3</sub> and PuCl<sub>3</sub>, which were parameterized in previous works, are also included. The results are discussed in the context of already existing theoretical and experimental datasets, showing good agreement with the literature. Predictions are extended to systems not considered in previous works. Notably, the results highlight the peculiarity of ThCl<sub>4</sub> compared to actinide trichlorides in terms of structural and thermophysical properties. The new force fields can be used in future works for the simulation of molten salts mixtures containing actinides.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155822"},"PeriodicalIF":2.8,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143854814","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

The role of pre-adsorbed hydrogen in facilitating water release at the Li4TiO4 surface: A first-principles study 预吸附氢在促进 Li4TiO4 表面水释放中的作用：第一原理研究

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

Journal of Nuclear Materials

Pub Date : 2025-04-11 DOI: 10.1016/j.jnucmat.2025.155823

Zhonghua Lu , Yanli Shi , Yuchen Liu , Huanhuan Liu , Xiuling Wang , Cong Zhang , Gaoyuan Wang , Jianqi Qi , Tiecheng Lu

Li₄TiO₄ is recognized as an attractive material for tritium breeding in fusion reactors, owing to its high lithium content. Understanding the process of surface water formation is important for optimizing its tritium release performance. In this work, we conducted first-principles calculations to investigate the transformation of desorbed H₂ to water molecules on the Li₄TiO₄ surface. This process encompasses the dissociative adsorption of H₂, the diffusion of hydrogen atoms, the abstraction of a surface oxygen atom to form water molecules, and their subsequent release. The possible adsorption sites, local diffusion pathways, and their corresponding activation energies are identified. We determined that the global energy barrier for the transformation from desorbed H₂ to desorbed H₂O in the pre-adsorbed hydrogen system is 1.62 eV, approximately 1 eV less than the 2.67 eV observed in the non-adsorbed hydrogen system. This reduction in the energy barrier suggests that pre-adsorbed hydrogen facilitates water release. The decrease in the energy barrier is attributed to the easier formation of oxygen vacancies on the surface in the presence of pre-adsorbed hydrogen. Our results are consistent with experimental observations that pre-adsorbed hydrogen promotes the release of tritiated water.

{"title":"The role of pre-adsorbed hydrogen in facilitating water release at the Li4TiO4 surface: A first-principles study","authors":"Zhonghua Lu , Yanli Shi , Yuchen Liu , Huanhuan Liu , Xiuling Wang , Cong Zhang , Gaoyuan Wang , Jianqi Qi , Tiecheng Lu","doi":"10.1016/j.jnucmat.2025.155823","DOIUrl":"10.1016/j.jnucmat.2025.155823","url":null,"abstract":"<div><div>Li<sub>4</sub>TiO<sub>4</sub> is recognized as an attractive material for tritium breeding in fusion reactors, owing to its high lithium content. Understanding the process of surface water formation is important for optimizing its tritium release performance. In this work, we conducted first-principles calculations to investigate the transformation of desorbed H<sub>2</sub> to water molecules on the Li<sub>4</sub>TiO<sub>4</sub> surface. This process encompasses the dissociative adsorption of H<sub>2</sub>, the diffusion of hydrogen atoms, the abstraction of a surface oxygen atom to form water molecules, and their subsequent release. The possible adsorption sites, local diffusion pathways, and their corresponding activation energies are identified. We determined that the global energy barrier for the transformation from desorbed H<sub>2</sub> to desorbed H<sub>2</sub>O in the pre-adsorbed hydrogen system is 1.62 eV, approximately 1 eV less than the 2.67 eV observed in the non-adsorbed hydrogen system. This reduction in the energy barrier suggests that pre-adsorbed hydrogen facilitates water release. The decrease in the energy barrier is attributed to the easier formation of oxygen vacancies on the surface in the presence of pre-adsorbed hydrogen. Our results are consistent with experimental observations that pre-adsorbed hydrogen promotes the release of tritiated water.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155823"},"PeriodicalIF":2.8,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143837760","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

Interpretation of dissolution behavior at the surface of uranium-zirconium oxide solid solutions

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

Journal of Nuclear Materials

Pub Date : 2025-04-10 DOI: 10.1016/j.jnucmat.2025.155820

Ryutaro Tonna , Takayuki Sasaki , Yoshihiro Okamoto , Taishi Kobayashi

The dissolution behavior of (U,Zr)O₂, the primary uranium solid phase in the fuel debris from the Fukushima Daiichi nuclear power plant accidents, was investigated thermodynamically and kinetically under atmospheric conditions. Cubic (U,Zr)O₂ samples with a uniform solid solution of Zr were prepared using wet chemistry methods, and static batch immersion tests were conducted. In strongly acidic conditions, where the solubility of U and Zr exceeded their concentrations, congruent dissolution of both elements was observed with (U,Zr)O₂ dissolving at the same rate as UO₂. In moderately acidic conditions, where the U solubility was higher than its concentration with Zr reaching a steady state at lower solubility, the U dissolution rate from (U,Zr)O₂ decreased compared to UO₂. In the presence of oxalic acid, with increased Zr solubility due to the formation of complexes, the U dissolution rate from (U,Zr)O₂ did not decrease. This indicates that Zr in (U,Zr)O₂ formed a secondary solid phase on the solid surface under conditions of lower Zr solubility, which in turn suppressed the oxidative dissolution of U.

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引用次数: 0

On the microstructural evolution and hydriding behavior of dilute Zr-2.5Nb-Y alloys

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

Journal of Nuclear Materials

Pub Date : 2025-04-10 DOI: 10.1016/j.jnucmat.2025.155819

Y. Pushpalatha Devi , K.V. Mani Krishna , N. Keskar , J.B. Singh , R.N. Singh

Zr-2.5Nb alloy is a critical pressure tube material in pressurized heavy water reactors (PHWRs), as its performance directly influences the operational life and safety of the reactor. However, the formation of hydrides in this alloy can detrimentally affect its mechanical properties. In the present study, yttrium was employed as a dilute addition to mitigate hydride embrittlement. Alloys were prepared with varying yttrium contents of 0.5 to 2 wt %. Yttrium addition resulted in the formation of fine yttria precipitates. The hot deformed microstructures of alloys with Y exhibited significant differences in morphology of the phases and prior β phase fraction when compared to reference Zr2.5Nb alloy. Dilatometry studies indicated that yttrium addition led to a reduction in the β-transus temperature of the alloy. The hydride behavior of the alloys was also examined, showing that yttrium significantly reduced hydride size to <20 μm, compared to a range of 10–150 μm in the absence of yttrium (Zr-2.5Nb). This comprehensive study of the microstructure and hydriding behavior, with the addition of yttrium to the Zr-2.5Nb alloy, suggests that yttrium may be considered for improving the alloy's performance in nuclear applications, in view of the mitigation of hydride embrittlement.

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