Pub Date : 2025-11-25DOI: 10.1016/j.jnucmat.2025.156334
Q. Barrès , O. Tissot , E. Meslin , I. Mouton , M. Loyer-Prost , C. Pareige
The 4th generation of nuclear reactors currently under study is based on fast neutrons. The high energy of neutrons induces new constraints for the surrounding structural materials such as the reactor vessel and the cladding. Materials capable of withstanding the new operating conditions must be found. Ferritic/Martensitic and ODS steels are good candidates for addressing corrosion, mechanical and irradiations issues. Under irradiation, the creation and migration of point defects leads to various mechanisms that modify the initial properties of materials. Radiation induced segregation (RIS) is one of these mechanisms. RIS will occur based on various parameters related to materials and the radiative environment like temperature or dose. This paper presents the quantification of RIS on different types of grain boundaries (GB) in an FeCr model alloy. Correlative analyses before and after irradiation have been conducted on the same GB structure using Electron Backscatter Diffraction (EBSD) and Transmission Kikuchi Diffraction (TKD) techniques. Chemical quantifications were performed using Atom Probe Tomography (APT). Systematic W-shape Cr profile across GB after irradiation is revealed. The extent of this phenomenon depends on the structure of the GB being studied.
{"title":"Coupled EBSD/TKD/APT study of segregation induced by irradiation in a Fe-13at.%Cr model alloy through different grain boundaries type","authors":"Q. Barrès , O. Tissot , E. Meslin , I. Mouton , M. Loyer-Prost , C. Pareige","doi":"10.1016/j.jnucmat.2025.156334","DOIUrl":"10.1016/j.jnucmat.2025.156334","url":null,"abstract":"<div><div>The 4th generation of nuclear reactors currently under study is based on fast neutrons. The high energy of neutrons induces new constraints for the surrounding structural materials such as the reactor vessel and the cladding. Materials capable of withstanding the new operating conditions must be found. Ferritic/Martensitic and ODS steels are good candidates for addressing corrosion, mechanical and irradiations issues. Under irradiation, the creation and migration of point defects leads to various mechanisms that modify the initial properties of materials. Radiation induced segregation (RIS) is one of these mechanisms. RIS will occur based on various parameters related to materials and the radiative environment like temperature or dose. This paper presents the quantification of RIS on different types of grain boundaries (GB) in an FeCr model alloy. Correlative analyses before and after irradiation have been conducted on the same GB structure using Electron Backscatter Diffraction (EBSD) and Transmission Kikuchi Diffraction (TKD) techniques. Chemical quantifications were performed using Atom Probe Tomography (APT). Systematic W-shape Cr profile across GB after irradiation is revealed. The extent of this phenomenon depends on the structure of the GB being studied.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156334"},"PeriodicalIF":3.2,"publicationDate":"2025-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145681605","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}
Pub Date : 2025-11-25DOI: 10.1016/j.jnucmat.2025.156333
Lawrence Coghlan , Robert Burrows , Ronald N. Clark , David Kumar , Mariia Zimina , Aya Shin , Jonathan Hawes , Tomas Martin
Alloy 800 is currently being considered as a suitable alloy for use within the next generation of High Temperature Gas Cooled Reactors (HTGRs). These reactors will operate using Helium as the heat transfer gas at temperatures of 700 °C and higher. HTGR materials need to be able to withstand these high temperatures over service lives of multiple decades in the presence of helium and impurities present either within the gas or from ingress to the system.
This work exposed Alloy 800 at 750 °C in a helium atmosphere for up to 1000 h and characterised the materials using advanced microscopy techniques. Within the bulk material, Mn and Cr segregation to the grain boundaries took place at <150 h of exposure due to the high temperature and similar segregation was seen towards the surface, both along grain boundaries and within grains. A non-uniform Cr- and Mn-rich oxide forms on the surface of Alloy 800 after 150 h with the uniformity and thickness increasing with exposure duration.
Al, Ti, Si and Mn are all seen to oxidise ahead of the bulk material and before Cr, with the development of an Al-rich internal oxidation zone and the formation of a non-uniform Mn-rich oxide forming at the surface of the alloy. Cr later migrates to the surface and forms part of this oxide. The thermodynamics of the oxidation reactions are in agreement with Ellingham diagram predictions with the alloying elements oxidising in order of stability.
{"title":"Corrosion of alloy 800 exposed to high temperature helium","authors":"Lawrence Coghlan , Robert Burrows , Ronald N. Clark , David Kumar , Mariia Zimina , Aya Shin , Jonathan Hawes , Tomas Martin","doi":"10.1016/j.jnucmat.2025.156333","DOIUrl":"10.1016/j.jnucmat.2025.156333","url":null,"abstract":"<div><div>Alloy 800 is currently being considered as a suitable alloy for use within the next generation of High Temperature Gas Cooled Reactors (HTGRs). These reactors will operate using Helium as the heat transfer gas at temperatures of 700 °C and higher. HTGR materials need to be able to withstand these high temperatures over service lives of multiple decades in the presence of helium and impurities present either within the gas or from ingress to the system.</div><div>This work exposed Alloy 800 at 750 °C in a helium atmosphere for up to 1000 h and characterised the materials using advanced microscopy techniques. Within the bulk material, Mn and Cr segregation to the grain boundaries took place at <150 h of exposure due to the high temperature and similar segregation was seen towards the surface, both along grain boundaries and within grains. A non-uniform Cr- and Mn-rich oxide forms on the surface of Alloy 800 after 150 h with the uniformity and thickness increasing with exposure duration.</div><div>Al, Ti, Si and Mn are all seen to oxidise ahead of the bulk material and before Cr, with the development of an Al-rich internal oxidation zone and the formation of a non-uniform Mn-rich oxide forming at the surface of the alloy. Cr later migrates to the surface and forms part of this oxide. The thermodynamics of the oxidation reactions are in agreement with Ellingham diagram predictions with the alloying elements oxidising in order of stability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156333"},"PeriodicalIF":3.2,"publicationDate":"2025-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145681607","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}
Pub Date : 2025-11-25DOI: 10.1016/j.jnucmat.2025.156315
Stefano Cipelli , Arshad Hussain , Giorgio Dilecce , Paolo Francesco Ambrico , Domenico Aceto , Anna Cremona , Irene Casiraghi , Olga De Pascale , Laura Laguardia , Matteo Pedroni , Daria Ricci , Dario Ripamonti , Jimmy Scionti , Andrea Uccello
Laser-Induced Breakdown Spectroscopy (LIBS) analyzes the optical emission of a laser-induced plasma to determine the elemental or isotopic composition of a material. In nuclear fusion research, LIBS is a promising diagnostic for remote monitoring of Plasma-Facing Components (PFCs). Accurate interpretation of LIBS measurements, especially for depth profiling and fuel retention studies, requires the knowledge of ablation rate and laser-induced thermal effects. This work presents a two-dimensional thermal model, developed using COMSOL Multiphysics, to simulate nanosecond laser ablation of tungsten, a fusion-relevant metal, and silicon, selected for its different thermophysical properties. The model solves the heat conduction equation considering temperature-dependent material properties, phase change, latent heat consumption, and plasma plume shielding included via an exponential attenuation factor. Material removal is implemented through a mesh deformation velocity dependent on the assumed ablation regime. Model validation was performed by comparing computations with laser ablation experiments in vacuum, using a 10 ns Nd:YAG laser. Ablation craters were characterized by optical microscopy, mechanical profilometry, and Scanning Electron Microscopy (SEM). For tungsten, normal evaporation was considered to describe ablation for pulse energy ( considering central crater); for silicon phase explosion dominated at (). The model reproduced ablated volume, depth, and crater diameter, obtaining relative discrepancies on depth resolution prediction around 20 %. These results demonstrate the potential of a physics-based model to predict LIBS crater features, supporting parameter optimization and aiding interpretation of depth profiling and fuel retention measurements.
{"title":"Nanosecond laser ablation modeling of silicon and tungsten as support activity for LIBS diagnostic","authors":"Stefano Cipelli , Arshad Hussain , Giorgio Dilecce , Paolo Francesco Ambrico , Domenico Aceto , Anna Cremona , Irene Casiraghi , Olga De Pascale , Laura Laguardia , Matteo Pedroni , Daria Ricci , Dario Ripamonti , Jimmy Scionti , Andrea Uccello","doi":"10.1016/j.jnucmat.2025.156315","DOIUrl":"10.1016/j.jnucmat.2025.156315","url":null,"abstract":"<div><div>Laser-Induced Breakdown Spectroscopy (LIBS) analyzes the optical emission of a laser-induced plasma to determine the elemental or isotopic composition of a material. In nuclear fusion research, LIBS is a promising diagnostic for remote monitoring of Plasma-Facing Components (PFCs). Accurate interpretation of LIBS measurements, especially for depth profiling and fuel retention studies, requires the knowledge of ablation rate and laser-induced thermal effects. This work presents a two-dimensional thermal model, developed using COMSOL Multiphysics, to simulate nanosecond laser ablation of tungsten, a fusion-relevant metal, and silicon, selected for its different thermophysical properties. The model solves the heat conduction equation considering temperature-dependent material properties, phase change, latent heat consumption, and plasma plume shielding included via an exponential attenuation factor. Material removal is implemented through a mesh deformation velocity dependent on the assumed ablation regime. Model validation was performed by comparing computations with laser ablation experiments in vacuum, using a 10 ns Nd:YAG laser. Ablation craters were characterized by optical microscopy, mechanical profilometry, and Scanning Electron Microscopy (SEM). For tungsten, normal evaporation was considered to describe ablation for <span><math><mrow><mn>0.4</mn><mo>−</mo><mn>2</mn><mspace></mspace><mtext>mJ</mtext></mrow></math></span> pulse energy (<span><math><mrow><mo>∼</mo><mn>4</mn><mo>−</mo><mn>20</mn><mspace></mspace><mi>J</mi><mo>/</mo><msup><mrow><mrow><mi>c</mi></mrow><mi>m</mi></mrow><mn>2</mn></msup></mrow></math></span> considering central crater); for silicon phase explosion dominated at <span><math><mrow><mn>4</mn><mo>−</mo><mn>20</mn><mspace></mspace><mtext>mJ</mtext></mrow></math></span> (<span><math><mrow><mo>∼</mo><mn>25</mn><mo>−</mo><mn>60</mn><mspace></mspace><mi>J</mi><mo>/</mo><msup><mrow><mrow><mi>c</mi></mrow><mi>m</mi></mrow><mn>2</mn></msup></mrow></math></span>). The model reproduced ablated volume, depth, and crater diameter, obtaining relative discrepancies on depth resolution prediction around 20 %. These results demonstrate the potential of a physics-based model to predict LIBS crater features, supporting parameter optimization and aiding interpretation of depth profiling and fuel retention measurements.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156315"},"PeriodicalIF":3.2,"publicationDate":"2025-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145681608","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}
In this work, single crystals of a new (UO2)0.76Th0.24(HCOO)2.46(H2O)0.76 compound were obtained. The phase composition was studied using X-ray powder diffraction, while the crystal structure was solved by single-crystal X-ray diffraction. The Th/(Th+U) ratios were determined using EDS (27.5 %), ICP-MS (25.9 %) and XRF (26.3 %). TGA was performed, confirming the water content in the compound and providing insight into its high-temperature decomposition behavior. The crystal structure of this new structural type has been solved for the first time with high precision (R = 0.02). The uranyl-thorium formate crystallizes in the orthorhombic system, space group Fddd, a = 0.5959(3), b = 1.9355(9), c= 1.1560(6) nm, V = 1.3332(11) nm3, Z = 8. Incorporation of thorium into the crystal structure of uranyl formate monohydrate, with a Th/(Th+U) ratio of up to 30 wt%, leads to structural disorder and a change in space group Fdd2→Fddd. The Th4+ and U6+cations are statistically disordered and occupy a single crystallographic site. The crystal structure can be described as the presence of two rotationally disordered frameworks consists of a uranyl pentagonal bipyramid coordinated by four formate groups. The topology in the crystal structure of (UO2)0.76Th0.24(HCOO)2.46(H2O)0.76 features eight-membered rings and corresponds to the cc2–1:2–2 topology type. Current work provides crucial structural insights into the formation mechanism of homogeneous actinide precursors that can enable the production of high-quality (U,Th)O₂-based MOX fuels through simplified, single-step denitration and decomposition processes.
本文获得了一种新的（UO2）0.76 th0.24 (HCOO)2.46(H2O)0.76化合物的单晶。用x射线粉末衍射研究了相组成，用x射线单晶衍射分析了晶体结构。采用EDS（27.5%）、ICP-MS（25.9%）和XRF（26.3%）测定Th/(Th+U)比值。进行TGA分析，确定了化合物中的水分含量，并对其高温分解行为进行了深入研究。这种新型结构类型的晶体结构首次得到高精度解析（R = 0.02）。甲酸铀酰钍在正交晶系中结晶，空间群Fddd， a = 0.5959(3), b = 1.9355(9), c = 1.1560(6) nm, V = 1.3332(11) nm3, Z = 8。在甲酸铀酰一水合物的晶体结构中加入钍，其Th/(Th+U)比高达30 wt%，导致结构紊乱，空间群Fdd2→Fddd发生变化。Th4+和U6+阳离子在统计上是无序的，占据一个单晶位。晶体结构可以描述为由四个甲酸基配位的铀酰五边形双金字塔组成的两个旋转无序框架的存在。（UO2）0.76 th0.24 (HCOO)2.46(H2O)0.76晶体结构的拓扑结构为八元环，对应于cc2-1:2-2拓扑类型。目前的工作为均相锕系前体的形成机制提供了关键的结构见解，这些前体可以通过简化的单步脱硝和分解过程生产高质量的（U,Th）O₂基MOX燃料。
{"title":"The crystal structure of novel solid solution (UO2)0.76Th0.24(HCOO)2.46(H2O)0.76","authors":"R.A. Serebryanskikh , S.V. Demina , A.S. Aloy , V.Yu. Grishaev , S.M. Aksenov","doi":"10.1016/j.jnucmat.2025.156335","DOIUrl":"10.1016/j.jnucmat.2025.156335","url":null,"abstract":"<div><div>In this work, single crystals of a new (UO<sub>2</sub>)<sub>0.76</sub>Th<sub>0.24</sub>(HCOO)<sub>2.46</sub>(H<sub>2</sub>O)<sub>0.76</sub> compound were obtained. The phase composition was studied using X-ray powder diffraction, while the crystal structure was solved by single-crystal X-ray diffraction. The Th/(Th+<em>U</em>) ratios were determined using EDS (27.5 %), ICP-MS (25.9 %) and XRF (26.3 %). TGA was performed, confirming the water content in the compound and providing insight into its high-temperature decomposition behavior. The crystal structure of this new structural type has been solved for the first time with high precision (<em>R</em> = 0.02). The uranyl-thorium formate crystallizes in the orthorhombic system, space group <em>Fddd, a</em> = 0.5959(3), <em>b</em> = 1.9355(9), <em>c</em> <em>=</em> 1.1560(6) nm, <em>V</em> = 1.3332(11) nm<sup>3</sup>, <em>Z</em> = 8. Incorporation of thorium into the crystal structure of uranyl formate monohydrate, with a Th/(Th+<em>U</em>) ratio of up to 30 wt%, leads to structural disorder and a change in space group <em>Fdd</em>2→<em>Fddd</em>. The Th<sup>4+</sup> and U<sup>6+</sup>cations are statistically disordered and occupy a single crystallographic site. The crystal structure can be described as the presence of two rotationally disordered frameworks consists of a uranyl pentagonal bipyramid coordinated by four formate groups. The topology in the crystal structure of (UO<sub>2</sub>)<sub>0.76</sub>Th<sub>0.24</sub>(HCOO)<sub>2.46</sub>(H<sub>2</sub>O)<sub>0.76</sub> features eight-membered rings and corresponds to the <em>cc</em>2–1:2–2 topology type. Current work provides crucial structural insights into the formation mechanism of homogeneous actinide precursors that can enable the production of high-quality (U,Th)O₂-based MOX fuels through simplified, single-step denitration and decomposition processes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156335"},"PeriodicalIF":3.2,"publicationDate":"2025-11-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145616232","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}
Pub Date : 2025-11-24DOI: 10.1016/j.jnucmat.2025.156325
Emile Mukiza , Quoc Tri Phung , Suresh Seetharam , Lander Frederickx , Ken Verguts , Eef Weetjens , Ziyou Yu , Rodrigo de Oliveira-Silva , Dimitrios Sakellariou , Geert De Schutter
This study examines the effect of gamma radiation on geopolymerization kinetics and early age microstructural development in metakaolin-based geopolymers incorporating simulated cesium (Cs) and strontium (Sr) radioactive waste. The effects of heat induced during gamma irradiation were decoupled from those of gamma ionizing radiation and assessed independently. Furthermore, a comparative analysis was conducted between the effects of gamma radiation from Co-60 source, commonly utilized in irradiation studies, and those from Cs-137 source associated with Cs- and Sr-containing waste. The results indicate that gamma radiation-induced heat only impacts the early reaction kinetics, without affecting the total 28-day degree of reaction. Co-60 irradiation does not interfere with the geopolymerization process, but Cs-137 exposure negatively affects the geopolymerization and subsequent formation of N–A–S–H gel. One of the evidences is the highest proportion of unreacted MK of 26.3 % with Cs-137 exposure compared to 18.6 % with Co-60 exposure at the same cumulative dose. This implies that the commonly used Co-60-based irradiation underestimates the radiation impact in real-world Cs/Sr waste immobilization. In addition, N–A–S–H gel exhibited radiation stability as indicated by FTIR, TGA, quantitative XRD, NMR, and SEM analysis. Based on the results, metakaolin-based geopolymers are promising as inexpensive alternatives to vitrification for the immobilization of Cs-137 and Sr-90-containing waste. Future studies using in-situ irradiation are needed to assess the effect of gamma radiation on the entire geopolymerization process.
{"title":"Effect of gamma radiation on geopolymerization and microstructure development of metakaolin-based geopolymer for Cesium and Strontium waste immobilization","authors":"Emile Mukiza , Quoc Tri Phung , Suresh Seetharam , Lander Frederickx , Ken Verguts , Eef Weetjens , Ziyou Yu , Rodrigo de Oliveira-Silva , Dimitrios Sakellariou , Geert De Schutter","doi":"10.1016/j.jnucmat.2025.156325","DOIUrl":"10.1016/j.jnucmat.2025.156325","url":null,"abstract":"<div><div>This study examines the effect of gamma radiation on geopolymerization kinetics and early age microstructural development in metakaolin-based geopolymers incorporating simulated cesium (Cs) and strontium (Sr) radioactive waste. The effects of heat induced during gamma irradiation were decoupled from those of gamma ionizing radiation and assessed independently. Furthermore, a comparative analysis was conducted between the effects of gamma radiation from Co-60 source, commonly utilized in irradiation studies, and those from Cs-137 source associated with Cs- and Sr-containing waste. The results indicate that gamma radiation-induced heat only impacts the early reaction kinetics, without affecting the total 28-day degree of reaction. Co-60 irradiation does not interfere with the geopolymerization process, but Cs-137 exposure negatively affects the geopolymerization and subsequent formation of N–A–S–H gel. One of the evidences is the highest proportion of unreacted MK of 26.3 % with Cs-137 exposure compared to 18.6 % with Co-60 exposure at the same cumulative dose. This implies that the commonly used Co-60-based irradiation underestimates the radiation impact in real-world Cs/Sr waste immobilization. In addition, N–A–S–H gel exhibited radiation stability as indicated by FTIR, TGA, quantitative XRD, NMR, and SEM analysis. Based on the results, metakaolin-based geopolymers are promising as inexpensive alternatives to vitrification for the immobilization of Cs-137 and Sr-90-containing waste. Future studies using in-situ irradiation are needed to assess the effect of gamma radiation on the entire geopolymerization process.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156325"},"PeriodicalIF":3.2,"publicationDate":"2025-11-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145616238","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}
Pub Date : 2025-11-24DOI: 10.1016/j.jnucmat.2025.156293
Yifan Li , Yaolin Guo , Zhen Liu , Bin Gu , Jingyu Zhang , Nianxiang Qiu , Zheyu Hu , Muhammad Adnan , Shiyu Du
This study presents a cluster dynamics model that integrates the nucleation of large clusters with their transformation into dislocation loops, enabling accurate prediction of irradiation-induced growth strain in zirconium across a broad range of pre-existing dislocation densities. The model captures the dynamic evolution of dislocation networks as sinks for point defects and mobile clusters and employs a multi-step parameter optimization. Results show that the unified model accurately reproduces strain evolution in both annealed (4.62 % error) and cold-worked (5.78 % error) zirconium alloys during the “breakaway” phase, the nonlinear accelerated growth stage at high displacements per atom (dpa). As dpa increases, specifically, the strain is successively governed by dislocation lines, interstitial loops and finally vacancy loops which drive the ”breakaway” stage. Notably, a critical dislocation density of is identified, above which a tenfold increase in dislocation density can amplify strain by 220 %, due to enhanced nucleation of vancancy loops. The study establishes a theoretical framework linking pre-existing substructure to irradiation-induced growth strain, providing a mechanistic basis for predicting anisotropic deformation of zirconium alloys under service conditions.
{"title":"Cluster dynamics simulation of irradiation-induced growth strain in zr alloys: Role of continuous pre-existing dislocation densities","authors":"Yifan Li , Yaolin Guo , Zhen Liu , Bin Gu , Jingyu Zhang , Nianxiang Qiu , Zheyu Hu , Muhammad Adnan , Shiyu Du","doi":"10.1016/j.jnucmat.2025.156293","DOIUrl":"10.1016/j.jnucmat.2025.156293","url":null,"abstract":"<div><div>This study presents a cluster dynamics model that integrates the nucleation of large clusters with their transformation into dislocation loops, enabling accurate prediction of irradiation-induced growth strain in zirconium across a broad range of pre-existing dislocation densities. The model captures the dynamic evolution of dislocation networks as sinks for point defects and mobile clusters and employs a multi-step parameter optimization. Results show that the unified model accurately reproduces strain evolution in both annealed (4.62 % error) and cold-worked (5.78 % error) zirconium alloys during the “breakaway” phase, the nonlinear accelerated growth stage at high displacements per atom (dpa). As dpa increases, specifically, the strain is successively governed by dislocation lines, interstitial loops and finally vacancy loops which drive the ”breakaway” stage. Notably, a critical dislocation density of <span><math><mrow><mn>1</mn><mo>×</mo><msup><mn>10</mn><mn>13</mn></msup><mspace></mspace><msup><mtext>m</mtext><mrow><mo>−</mo><mn>2</mn></mrow></msup></mrow></math></span> is identified, above which a tenfold increase in dislocation density can amplify strain by 220 %, due to enhanced nucleation of vancancy loops. The study establishes a theoretical framework linking pre-existing substructure to irradiation-induced growth strain, providing a mechanistic basis for predicting anisotropic deformation of zirconium alloys under service conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156293"},"PeriodicalIF":3.2,"publicationDate":"2025-11-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145733473","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}
Pub Date : 2025-11-23DOI: 10.1016/j.jnucmat.2025.156323
Fengjiao Ye , Haibiao Wu , Peng Zhang , Xingzhong Cao , Peng Kuang , Fuyan Liu , Runsheng Yu , Te Zhu , Baoyi Wang
The behavior of hydrogen (H) and its interaction with defects in a Fe6Cr1.2Mn0.8Cu1.5Mo0.5 alloy irradiated with 30 keV H ions under various fluence and temperatures was investigated using positron annihilation spectroscopy. Results reveal that significant numbers of vacancy-type defects are formed in samples irradiated at room temperature, where H atoms combine with vacancies to form H-vacancy complexes (HmVn). The irradiation damage profile shows substantial variation with depth. In high-fluence irradiated samples, the formation of HmVn complexes (m > n) suppresses the increase in effective open volume defects. The vacancy concentration gradually decreases with increasing irradiation temperature. Coincidence Doppler Broadening (CDB) spectroscopy demonstrates that vacancies generated in alloys irradiated at 150 °C exhibit migration capability, ultimately aggregating into vacancy clusters. Notably, Cu precipitation sites are identified as preferential nucleation points for vacancy cluster formation. Elevated irradiation temperatures lead to a gradual reduction in vacancy cluster size and accelerated recovery processes. At 450 °C irradiation temperature, vacancy defects dissociated from Cu precipitates undergo nearly complete recovery. This study establishes positron annihilation spectroscopy as an effective methodology for elucidating H behavior and HmVn complex evolution in multi-principal element alloys.
{"title":"Study on the interaction between hydrogen and irradiation defects in iron-based multi-principal element alloy using positron annihilation spectroscopy","authors":"Fengjiao Ye , Haibiao Wu , Peng Zhang , Xingzhong Cao , Peng Kuang , Fuyan Liu , Runsheng Yu , Te Zhu , Baoyi Wang","doi":"10.1016/j.jnucmat.2025.156323","DOIUrl":"10.1016/j.jnucmat.2025.156323","url":null,"abstract":"<div><div>The behavior of hydrogen (H) and its interaction with defects in a Fe<sub>6</sub>Cr<sub>1.2</sub>Mn<sub>0.8</sub>Cu<sub>1.5</sub>Mo<sub>0.5</sub> alloy irradiated with 30 keV H ions under various fluence and temperatures was investigated using positron annihilation spectroscopy. Results reveal that significant numbers of vacancy-type defects are formed in samples irradiated at room temperature, where H atoms combine with vacancies to form H-vacancy complexes (HmVn). The irradiation damage profile shows substantial variation with depth. In high-fluence irradiated samples, the formation of HmVn complexes (<em>m</em> > <em>n</em>) suppresses the increase in effective open volume defects. The vacancy concentration gradually decreases with increasing irradiation temperature. Coincidence Doppler Broadening (CDB) spectroscopy demonstrates that vacancies generated in alloys irradiated at 150 °C exhibit migration capability, ultimately aggregating into vacancy clusters. Notably, Cu precipitation sites are identified as preferential nucleation points for vacancy cluster formation. Elevated irradiation temperatures lead to a gradual reduction in vacancy cluster size and accelerated recovery processes. At 450 °C irradiation temperature, vacancy defects dissociated from Cu precipitates undergo nearly complete recovery. This study establishes positron annihilation spectroscopy as an effective methodology for elucidating H behavior and HmVn complex evolution in multi-principal element alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156323"},"PeriodicalIF":3.2,"publicationDate":"2025-11-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145616219","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}
Pub Date : 2025-11-22DOI: 10.1016/j.jnucmat.2025.156324
Jun Wang, Bo Zhang, Lixia Zhang, Qing Chang, Zhan Sun, Ming Gao
Conventional hot pressure diffusion bonding (HPDB) of Zr-4 alloy has been conducted at relatively high temperatures (over 800 °C). In HPDB processing, the entire base material is uniformly heated, which causes excessive deformation and grain coarsening. In this study, a novel approach was proposed by applying an electric current diffusion bonding (ECDB) method. This novel approach can address the challenges given by HPDB and achieve a low-deformation and high-strength bonding simultaneously. The shear strength of the ECDB joint reaches 369 MPa under 650 °C / 40 min / 20 MPa, which is 92 % of that of the as-received Zr-4. The temperature is over 150 °C lower than that of the HPDB to achieve a comparable joint strength. Additionally, the joint deformation rate of the ECDB was 34 % lower than that of the HPDB. The base metal microstructure of the ECDB exhibited a smaller average grain size, as high temperature only exists in a microscopic region at the interface (≈1.4 μm). Under the same parameters, the electric current realized complete bonding of the diffusion bonding joints, and no obvious interface was observable.
传统的Zr-4合金热压扩散键合（HPDB）是在相对较高的温度（800℃以上）下进行的。在HPDB加工中，整个基材被均匀加热，导致过度变形和晶粒粗化。本研究提出了一种应用电流扩散键合（ECDB）的新方法。这种新方法可以解决HPDB所面临的挑战，同时实现低变形和高强度的粘合。在650℃/ 40 min / 20 MPa条件下，ECDB接头抗剪强度达到369 MPa，达到Zr-4的92%。温度比HPDB低150°C以上，以达到相当的接头强度。此外，ECDB的关节变形率比HPDB低34%。由于高温仅存在于界面处的微观区域（≈1.4 μm）， ECDB的基体金属微观结构表现出较小的平均晶粒尺寸。在相同的参数下，电流实现了扩散焊接头的完全结合，没有观察到明显的界面。
{"title":"Electric current induced low-temperature diffusion bonding mechanism of Zr-4 alloy","authors":"Jun Wang, Bo Zhang, Lixia Zhang, Qing Chang, Zhan Sun, Ming Gao","doi":"10.1016/j.jnucmat.2025.156324","DOIUrl":"10.1016/j.jnucmat.2025.156324","url":null,"abstract":"<div><div>Conventional hot pressure diffusion bonding (HPDB) of Zr-4 alloy has been conducted at relatively high temperatures (over 800 °C). In HPDB processing, the entire base material is uniformly heated, which causes excessive deformation and grain coarsening. In this study, a novel approach was proposed by applying an electric current diffusion bonding (ECDB) method. This novel approach can address the challenges given by HPDB and achieve a low-deformation and high-strength bonding simultaneously. The shear strength of the ECDB joint reaches 369 MPa under 650 °C / 40 min / 20 MPa, which is 92 % of that of the as-received Zr-4. The temperature is over 150 °C lower than that of the HPDB to achieve a comparable joint strength. Additionally, the joint deformation rate of the ECDB was 34 % lower than that of the HPDB. The base metal microstructure of the ECDB exhibited a smaller average grain size, as high temperature only exists in a microscopic region at the interface (≈1.4 μm). Under the same parameters, the electric current realized complete bonding of the diffusion bonding joints, and no obvious interface was observable.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156324"},"PeriodicalIF":3.2,"publicationDate":"2025-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145616239","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}
Pub Date : 2025-11-22DOI: 10.1016/j.jnucmat.2025.156321
Jian Chen , Jonathan Tatman , Zhili Feng , Roger Miller , Greg Frederick
Welding repair of irradiated nickel-based alloys, such as Alloy 182, poses a significant challenge due to helium-induced cracking (HeIC) and grain boundary degradation (GBD) in the heat affected zone, driven by helium accumulation at grain boundaries and welding-induced tensile stresses. This study investigates the weldability of irradiated Alloy 182 up to 15 wppm doped boron using both conventional laser welding and the Auxiliary Beam Stress Improved (ABSI) laser welding technique. While HeIC was observed at the weld toe of the entry pass in both methods due to the higher effective heat input associated with the initial pass directly on the base metal, no additional cracking occurred elsewhere, even at elevated helium concentrations. Optical and scanning electron microscopy analysis revealed that the ABSI technique, which introduces additional compressive stresses to counteract solidification-induced tensile stresses, significantly reduced GBD formation, lowering its total count from 1,230 to 339 and decreasing both average and maximum GBD lengths. These results demonstrate that the ABSI technique is a promising approach to mitigate helium-induced damage and improve the weldability of irradiated Alloy 182, offering a viable solution for structural repairs for long-term operation of existing nuclear reactors.
{"title":"Laser repair welding of irradiated alloy 182","authors":"Jian Chen , Jonathan Tatman , Zhili Feng , Roger Miller , Greg Frederick","doi":"10.1016/j.jnucmat.2025.156321","DOIUrl":"10.1016/j.jnucmat.2025.156321","url":null,"abstract":"<div><div>Welding repair of irradiated nickel-based alloys, such as Alloy 182, poses a significant challenge due to helium-induced cracking (HeIC) and grain boundary degradation (GBD) in the heat affected zone, driven by helium accumulation at grain boundaries and welding-induced tensile stresses. This study investigates the weldability of irradiated Alloy 182 up to 15 wppm doped boron using both conventional laser welding and the Auxiliary Beam Stress Improved (ABSI) laser welding technique. While HeIC was observed at the weld toe of the entry pass in both methods due to the higher effective heat input associated with the initial pass directly on the base metal, no additional cracking occurred elsewhere, even at elevated helium concentrations. Optical and scanning electron microscopy analysis revealed that the ABSI technique, which introduces additional compressive stresses to counteract solidification-induced tensile stresses, significantly reduced GBD formation, lowering its total count from 1,230 to 339 and decreasing both average and maximum GBD lengths. These results demonstrate that the ABSI technique is a promising approach to mitigate helium-induced damage and improve the weldability of irradiated Alloy 182, offering a viable solution for structural repairs for long-term operation of existing nuclear reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156321"},"PeriodicalIF":3.2,"publicationDate":"2025-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145616235","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}
Pub Date : 2025-11-22DOI: 10.1016/j.jnucmat.2025.156322
Shuxian Sun , Pengbo Zhang , Mingliang Wei , Tingting Zou , Yaxia Wei , Pengfei Zheng
The formation, migration, and clustering behaviors of vacancy defects in different local environments on them in V-4Cr-4Ti alloy are investigated by first-principles calculations. Energetically, vacancy formation energy ranges from 1.54 to 2.64 eV in 67 different chemical environments, its mean value of 2.22 eV is lower than that of pure V. The influence of Cr and Ti on vacancies mainly is a short-range effect, and Ti reduces vacancy formation energy while Cr increases it. For complex defects, the di- and tri-vacancy formation energies range from 3.22∼4.74 eV and 5.15∼6.93 eV, respectively. The lower formation energies occur in Ti-rich regions. Vacancy clusters tend to form compact structures in Ti regions with the larger binding energies of 1.57∼2.23 eV. The stability of vacancyn (Vacn) clusters increases with the size, the Ti-Vacn clusters are more stable than Vacn and Cr-Vacn. Kinetically, the migration property and the clustering process of vacancies are predicted by the climbing image nudged elastic band (CI-NEB) method. Local environments have a large effect on vacancy migration path and barrier, vacancy diffusing towards Ti shows a lower barrier of 0.24 eV than 0.45 eV in pure V. Vacancies and Ti-vacancy easily migrate to form Ti-vacancy clusters with higher stability. Vacancy diffusing toward Cr shows a higher barrier of 0.86 eV, the migration and clustering of vacancies are hindered in Cr regions. These findings presented that Ti can stabilize vacancies and small vacancy clusters in the V-4Cr-4Ti alloy, Cr can alter vacancy migration paths and promote their movement toward Ti, leading to forming numerous of stable small-sized Ti-vacancy complexes, effectively suppressing the formation of void under irradiation.
{"title":"The formation, migration, and clustering behaviors of vacancy defects in different local environments of V-4Cr-4Ti alloy","authors":"Shuxian Sun , Pengbo Zhang , Mingliang Wei , Tingting Zou , Yaxia Wei , Pengfei Zheng","doi":"10.1016/j.jnucmat.2025.156322","DOIUrl":"10.1016/j.jnucmat.2025.156322","url":null,"abstract":"<div><div>The formation, migration, and clustering behaviors of vacancy defects in different local environments on them in V-4Cr-4Ti alloy are investigated by first-principles calculations. Energetically, vacancy formation energy ranges from 1.54 to 2.64 eV in 67 different chemical environments, its mean value of 2.22 eV is lower than that of pure V. The influence of Cr and Ti on vacancies mainly is a short-range effect, and Ti reduces vacancy formation energy while Cr increases it. For complex defects, the di- and tri-vacancy formation energies range from 3.22∼4.74 eV and 5.15∼6.93 eV, respectively. The lower formation energies occur in Ti-rich regions. Vacancy clusters tend to form compact structures in Ti regions with the larger binding energies of 1.57∼2.23 eV. The stability of vacancy<em><sub>n</sub></em> (Vac<em><sub>n</sub></em>) clusters increases with the size, the Ti-Vac<em><sub>n</sub></em> clusters are more stable than Vac<em><sub>n</sub></em> and Cr-Vac<em><sub>n</sub></em>. Kinetically, the migration property and the clustering process of vacancies are predicted by the climbing image nudged elastic band (CI-NEB) method. Local environments have a large effect on vacancy migration path and barrier, vacancy diffusing towards Ti shows a lower barrier of 0.24 eV than 0.45 eV in pure V. Vacancies and Ti-vacancy easily migrate to form Ti-vacancy clusters with higher stability. Vacancy diffusing toward Cr shows a higher barrier of 0.86 eV, the migration and clustering of vacancies are hindered in Cr regions. These findings presented that Ti can stabilize vacancies and small vacancy clusters in the V-4Cr-4Ti alloy, Cr can alter vacancy migration paths and promote their movement toward Ti, leading to forming numerous of stable small-sized Ti-vacancy complexes, effectively suppressing the formation of void under irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"620 ","pages":"Article 156322"},"PeriodicalIF":3.2,"publicationDate":"2025-11-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145616093","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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