Pub Date : 2025-12-29DOI: 10.1016/j.jnucmat.2025.156419
Xiang Li , Jinsong Zhang , Shuang Dai , Ke Wang , Yi Wang , Jia Tang , Shubo Yang , Qi Cao
In zirconium-alloy corrosion models, O2− movement through the zirconium oxide film is commonly considered as the rate-determining step. SEM-EDS results in this study indicate that O2− movement is driven by the built-in electric field rather than by concentration gradients. Molecular dynamics simulations were employed to investigate O2− mobility in several zirconium oxide structures, showing that monoclinic ZrO2 with a vertical grain boundary provides the highest migration ability. To investigate irradiation effects, the study focuses on low-dose displacement damage resulting from primary knock-on atom events. The simulations including mean square displacement, trajectory and diffusion coefficient results demonstrate that such low dose irradiation induced damage generally causes minimal change in O2− mobility, suggesting a limited influence on corrosion under these conditions.
{"title":"Corrosion study of zirconium alloys via molecular dynamics simulations: I. oxygen migration in zirconium oxides under low-dose irradiation induced displacement damage","authors":"Xiang Li , Jinsong Zhang , Shuang Dai , Ke Wang , Yi Wang , Jia Tang , Shubo Yang , Qi Cao","doi":"10.1016/j.jnucmat.2025.156419","DOIUrl":"10.1016/j.jnucmat.2025.156419","url":null,"abstract":"<div><div>In zirconium-alloy corrosion models, <em>O<sup>2−</sup></em> movement through the zirconium oxide film is commonly considered as the rate-determining step. SEM-EDS results in this study indicate that <em>O<sup>2−</sup></em> movement is driven by the built-in electric field rather than by concentration gradients. Molecular dynamics simulations were employed to investigate <em>O<sup>2−</sup></em> mobility in several zirconium oxide structures, showing that monoclinic ZrO<sub>2</sub> with a vertical grain boundary provides the highest migration ability. To investigate irradiation effects, the study focuses on low-dose displacement damage resulting from primary knock-on atom events. The simulations including mean square displacement, trajectory and diffusion coefficient results demonstrate that such low dose irradiation induced damage generally causes minimal change in <em>O<sup>2−</sup></em> mobility, suggesting a limited influence on corrosion under these conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156419"},"PeriodicalIF":3.2,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882264","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 : 2025-12-29DOI: 10.1016/j.jnucmat.2025.156420
Yu Yang , Meijiao Huang , Jixue Sui , An Li , Guangming Shen , Xiaoyong Wu , Lu Wu , Mingzhang Lin
The concentrations of soluble metal ions in the primary water circulation loop of pressurized water reactors can greatly affect the deposition behaviors of corrosion products on zirconium alloy cladding tubes at high temperatures, which is referred to as CRUD. However, while considerable progress has been made in the development of CRUD deposition models, the applicability of these models remains limited, and the impacts of Fe3+, Ni2+, Cr3+, Co2+, and Mn2+ on the microstructure of CRUD and the corrosion mechanism have not been fully incorporated into model development. The present work addresses these issues by combining experimental characterization with thermodynamic calculations to analyze the mechanism by which high-temperature aqueous solutions with different Fe3+, Ni2+, Cr3+, Co2+, and Mn2+ ion concentrations influence the CRUD deposition process on zirconium alloy cladding tubes. In addition to the composition and structure of CRUD deposits, the oxide film properties of the cladding tubes are further examined under subcooled nucleate boiling (SNB) and non-SNB conditions. The results indicate that increases in the weight of zirconium alloy cladding tubes under the non-SNB condition are due primarily to oxide film thickening under low metal ion concentrations. At high ion concentrations, a thinner oxide film was observed, which is likely attributed to modifications in the local chemical environment at the metal/oxide interface induced by the CRUD deposits. In contrast, the CRUD morphology changed under the SNB condition from cluster-like deposits without chimneys to porous structures with chimneys with increasing metal ion concentrations. Experimental results obtained under the SNB condition, in addition to the results of thermodynamic calculations of the Gibbs free energy, demonstrate that Fe3O4 and spinel compounds, such as CoFe2O4 and FeCr2O4, are the dominant phases within the CRUD deposits.
{"title":"Influence of multi-ion concentrations on corrosion product deposition on zirconium alloy cladding in high-temperature water","authors":"Yu Yang , Meijiao Huang , Jixue Sui , An Li , Guangming Shen , Xiaoyong Wu , Lu Wu , Mingzhang Lin","doi":"10.1016/j.jnucmat.2025.156420","DOIUrl":"10.1016/j.jnucmat.2025.156420","url":null,"abstract":"<div><div>The concentrations of soluble metal ions in the primary water circulation loop of pressurized water reactors can greatly affect the deposition behaviors of corrosion products on zirconium alloy cladding tubes at high temperatures, which is referred to as CRUD. However, while considerable progress has been made in the development of CRUD deposition models, the applicability of these models remains limited, and the impacts of Fe<sup>3+</sup>, Ni<sup>2+</sup>, Cr<sup>3+</sup>, Co<sup>2+</sup>, and Mn<sup>2+</sup> on the microstructure of CRUD and the corrosion mechanism have not been fully incorporated into model development. The present work addresses these issues by combining experimental characterization with thermodynamic calculations to analyze the mechanism by which high-temperature aqueous solutions with different Fe<sup>3+</sup>, Ni<sup>2+</sup>, Cr<sup>3+</sup>, Co<sup>2+</sup>, and Mn<sup>2+</sup> ion concentrations influence the CRUD deposition process on zirconium alloy cladding tubes. In addition to the composition and structure of CRUD deposits, the oxide film properties of the cladding tubes are further examined under subcooled nucleate boiling (SNB) and non-SNB conditions. The results indicate that increases in the weight of zirconium alloy cladding tubes under the non-SNB condition are due primarily to oxide film thickening under low metal ion concentrations. At high ion concentrations, a thinner oxide film was observed, which is likely attributed to modifications in the local chemical environment at the metal/oxide interface induced by the CRUD deposits. In contrast, the CRUD morphology changed under the SNB condition from cluster-like deposits without chimneys to porous structures with chimneys with increasing metal ion concentrations. Experimental results obtained under the SNB condition, in addition to the results of thermodynamic calculations of the Gibbs free energy, demonstrate that Fe<sub>3</sub>O<sub>4</sub> and spinel compounds, such as CoFe<sub>2</sub>O<sub>4</sub> and FeCr<sub>2</sub>O<sub>4</sub>, are the dominant phases within the CRUD deposits.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156420"},"PeriodicalIF":3.2,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882262","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 : 2025-12-29DOI: 10.1016/j.jnucmat.2025.156418
Diallo Barnes, Courtney Eckley, Mary Cernyar, Michael F. Simpson
To develop a process for chlorinating spent nuclear fuel (SNF) from light water reactors and dissolving it into a molten salt, the reactivity of several surrogate fission products (SFPs) with NiCl2 and ZrCl4 in molten LiCl – KCl was tested. ZrCl4 was created in-situ by reacting NiCl2 and Zr metal in the molten salt. Powders of SrO, La2O3, CeO2, Cs2O, and a Mo metal rod were all immersed in eutectic LiCl – KCl with initially 8.9 wt% NiCl2, likely exceeding the solubility limit. Salt samples were collected at several time intervals over 24 h and analyzed via inductively coupled plasma mass spectrometry (ICP – MS). NiCl2 did not directly react with any of the SFPs, but in situ formed ZrCl4 was reactive towards all the oxides. The Mo metal rod was virtually inert towards reaction with both NiCl2 and ZrCl4. Experiments were run at 773 or 873 K with or without stirring condition. Conversion of SrO, CeO2, and La2O3 to soluble chlorides ranged from 87 – 93 %, while Cs2O conversion was only 64 % on average. Some of the Ni metal by-product plated onto the Mo rod, which was unreactive towards chlorination. A solid containing Ni and ZrO2 settled to the bottom of the crucible.
{"title":"In – situ chlorination of simulated fission products from light water reactor fuel using ZrCl4","authors":"Diallo Barnes, Courtney Eckley, Mary Cernyar, Michael F. Simpson","doi":"10.1016/j.jnucmat.2025.156418","DOIUrl":"10.1016/j.jnucmat.2025.156418","url":null,"abstract":"<div><div>To develop a process for chlorinating spent nuclear fuel (SNF) from light water reactors and dissolving it into a molten salt, the reactivity of several surrogate fission products (SFPs) with NiCl<sub>2</sub> and ZrCl<sub>4</sub> in molten LiCl – KCl was tested. ZrCl<sub>4</sub> was created <em>in-situ</em> by reacting NiCl<sub>2</sub> and Zr metal in the molten salt. Powders of SrO, La<sub>2</sub>O<sub>3</sub>, CeO<sub>2</sub>, Cs<sub>2</sub>O, and a Mo metal rod were all immersed in eutectic LiCl – KCl with initially 8.9 wt% NiCl<sub>2</sub>, likely exceeding the solubility limit. Salt samples were collected at several time intervals over 24 h and analyzed via inductively coupled plasma mass spectrometry (ICP – MS). NiCl<sub>2</sub> did not directly react with any of the SFPs, but <em>in situ</em> formed ZrCl<sub>4</sub> was reactive towards all the oxides. The Mo metal rod was virtually inert towards reaction with both NiCl<sub>2</sub> and ZrCl<sub>4</sub>. Experiments were run at 773 or 873 K with or without stirring condition. Conversion of SrO, CeO<sub>2</sub>, and La<sub>2</sub>O<sub>3</sub> to soluble chlorides ranged from 87 – 93 %, while Cs<sub>2</sub>O conversion was only 64 % on average. Some of the Ni metal by-product plated onto the Mo rod, which was unreactive towards chlorination. A solid containing Ni and ZrO<sub>2</sub> settled to the bottom of the crucible.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156418"},"PeriodicalIF":3.2,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922725","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 : 2025-12-29DOI: 10.1016/j.jnucmat.2025.156422
Haiyan Liao , Xiaohan Deng , Weijiu Huang , Haibo Ruan , Shuai Lyu , Yuan Niu , Xiangkong Xu , Yongyao Su , Junjun Wang
This study systematically investigates the effects of Al content (3, 5, and 7 wt%) and a Cr-N diffusion barrier on the high-temperature steam oxidation performance of magnetron-sputtered FeCrAl coatings deposited on Zry-4. The tests were conducted at 1200 °C under simulated loss of coolant accident (LOCA) conditions in a steam environment. Results reveal that a minimum Al content of 5 wt% is required to form a continuous and dense Al2O3 layer, which significantly improves oxidation resistance. The Cr-N interlayer effectively suppressed the interdiffusion of Zr and Fe by promoting the in-situ formation of a continuous Zr(Fe,Cr)2 Laves phase barrier during oxidation. This barrier inhibited the development of low-melting-point Fe-Zr eutectic phases, thereby enhancing interfacial stability. Compared with a pure Cr interlayer, the Cr-N barrier demonstrated superior performance in mitigating interdiffusion and improving coating durability. These findings highlight the critical roles of Al content and the Cr-N interlayer in optimizing FeCrAl coatings for accident-tolerant fuel (ATF) claddings, offering essential insights for designing zirconium alloy coatings with superior oxidation resistance and reliability under extreme conditions.
{"title":"Effect of Al content and a Cr-N diffusion barrier on the high-temperature steam oxidation of FeCrAl coatings on Zry-4","authors":"Haiyan Liao , Xiaohan Deng , Weijiu Huang , Haibo Ruan , Shuai Lyu , Yuan Niu , Xiangkong Xu , Yongyao Su , Junjun Wang","doi":"10.1016/j.jnucmat.2025.156422","DOIUrl":"10.1016/j.jnucmat.2025.156422","url":null,"abstract":"<div><div>This study systematically investigates the effects of Al content (3, 5, and 7 wt%) and a Cr-N diffusion barrier on the high-temperature steam oxidation performance of magnetron-sputtered FeCrAl coatings deposited on Zry-4. The tests were conducted at 1200 °C under simulated loss of coolant accident (LOCA) conditions in a steam environment. Results reveal that a minimum Al content of 5 wt% is required to form a continuous and dense Al<sub>2</sub>O<sub>3</sub> layer, which significantly improves oxidation resistance. The Cr-N interlayer effectively suppressed the interdiffusion of Zr and Fe by promoting the in-situ formation of a continuous Zr(Fe,Cr)<sub>2</sub> Laves phase barrier during oxidation. This barrier inhibited the development of low-melting-point Fe-Zr eutectic phases, thereby enhancing interfacial stability. Compared with a pure Cr interlayer, the Cr-N barrier demonstrated superior performance in mitigating interdiffusion and improving coating durability. These findings highlight the critical roles of Al content and the Cr-N interlayer in optimizing FeCrAl coatings for accident-tolerant fuel (ATF) claddings, offering essential insights for designing zirconium alloy coatings with superior oxidation resistance and reliability under extreme conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156422"},"PeriodicalIF":3.2,"publicationDate":"2025-12-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882269","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 : 2025-12-28DOI: 10.1016/j.jnucmat.2025.156421
Jianguo Ma , Zhihong Liu , Chunwei Ma , Wei Wen , Huapeng Wu , Haibiao Ji , Rui Wang , Yuquan Kuang , Wangqi Shi , Haiying Xu , Weiping Fang , Zhiyong Wang , Yetao He
This paper systematically investigates the critical challenges in metal additive manufacturing (AM) technology for producing tungsten first wall components in nuclear fusion reactors—namely, residual stress and microcrack control. Research indicates that tungsten's high melting point (3422 °C) combined with AM's rapid cooling characteristics (10³-10⁴ K/s) results in melt pool temperature gradients exceeding 10⁶ K/m, inducing significant residual stresses. while the material exhibits a sharp decline in plasticity within the brittle-to-ductile transition temperature range of 450–650 K, making residual tensile stresses (>100 MPa) prone to triggering microcracks. Synergistic regulation through optimized scanning strategies (e.g., adjusting scan paths and energy distribution) and post-processing techniques achieves the following: Surface strengthening techniques (e.g., laser shock peening) form a 100–300 μm compressive stress layer, significantly reducing microcrack density; High-temperature heat treatment (1200–1400 °C) promotes crack healing through grain boundary migration. Data-driven machine learning methods enable real-time residual stress prediction, supporting intelligent process optimization. Future efforts should integrate multiscale simulations with irradiation experiments to validate and advance the engineering application of tungsten components in fusion reactors.
{"title":"Research progress on residual stress and microcrack control of tungsten fabricated via additive manufacturing","authors":"Jianguo Ma , Zhihong Liu , Chunwei Ma , Wei Wen , Huapeng Wu , Haibiao Ji , Rui Wang , Yuquan Kuang , Wangqi Shi , Haiying Xu , Weiping Fang , Zhiyong Wang , Yetao He","doi":"10.1016/j.jnucmat.2025.156421","DOIUrl":"10.1016/j.jnucmat.2025.156421","url":null,"abstract":"<div><div>This paper systematically investigates the critical challenges in metal additive manufacturing (AM) technology for producing tungsten first wall components in nuclear fusion reactors—namely, residual stress and microcrack control. Research indicates that tungsten's high melting point (3422 °C) combined with AM's rapid cooling characteristics (10³-10⁴ K/s) results in melt pool temperature gradients exceeding 10⁶ K/m, inducing significant residual stresses. while the material exhibits a sharp decline in plasticity within the brittle-to-ductile transition temperature range of 450–650 K, making residual tensile stresses (>100 MPa) prone to triggering microcracks. Synergistic regulation through optimized scanning strategies (e.g., adjusting scan paths and energy distribution) and post-processing techniques achieves the following: Surface strengthening techniques (e.g., laser shock peening) form a 100–300 μm compressive stress layer, significantly reducing microcrack density; High-temperature heat treatment (1200–1400 °C) promotes crack healing through grain boundary migration. Data-driven machine learning methods enable real-time residual stress prediction, supporting intelligent process optimization. Future efforts should integrate multiscale simulations with irradiation experiments to validate and advance the engineering application of tungsten components in fusion reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156421"},"PeriodicalIF":3.2,"publicationDate":"2025-12-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922654","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 : 2025-12-28DOI: 10.1016/j.jnucmat.2025.156417
Chong Liu , Dazhao Cheng , Jiahui Qu , Dehui Li , Yan Zhao , Jing Zhang
The evolution of helium bubbles in nickel-based alloys poses significant challenges to the structural integrity of nuclear reactor components. This study investigates the spatial heterogeneity and anisotropic evolution of helium bubbles in Inconel 617 under He ion irradiation (1 MeV, 3 × 10¹⁶ ions/cm², 830 °C) using a coupled experimental and anisotropic phase-field approach. Transmission electron microscopy reveals distinct bubble characteristics: the (Ni,Cr)O oxide/matrix interface exhibits the highest bubble density, while both the interface and the oxide interior share similar, fine bubble sizes (∼2.4 nm). In contrast, the Ni matrix hosts larger bubbles (∼3.8 nm) at a lower density. The anisotropic phase-field model, developed to quantitatively interpret these observations, identifies the coherent oxide interface as a potent defect sink. Its strong absorption of helium and vacancies promotes dense bubble nucleation while inducing a local vacancy depletion that suppresses coarsening in the oxide. Conversely, in the matrix, long-range diffusion enables the growth of sparse, large bubbles. Critically, the simulation reproduces the experimentally observed quadrilateral bubble morphology in the matrix, confirming the dominance of surface energy anisotropy in shaping bubbles. This work provides the first quantitative interpretation of anisotropic bubble evolution at oxide interfaces in IN617, highlighting the critical interplay between defect sink strength and crystallographic anisotropy. The coupled methodology establishes a predictive framework for designing radiation-tolerant microstructures.
{"title":"Spatially heterogeneous evolution of helium bubbles in He-irradiated Inconel 617: Experimental observation and anisotropic phase-field simulation","authors":"Chong Liu , Dazhao Cheng , Jiahui Qu , Dehui Li , Yan Zhao , Jing Zhang","doi":"10.1016/j.jnucmat.2025.156417","DOIUrl":"10.1016/j.jnucmat.2025.156417","url":null,"abstract":"<div><div>The evolution of helium bubbles in nickel-based alloys poses significant challenges to the structural integrity of nuclear reactor components. This study investigates the spatial heterogeneity and anisotropic evolution of helium bubbles in Inconel 617 under He ion irradiation (1 MeV, 3 × 10¹⁶ ions/cm², 830 °C) using a coupled experimental and anisotropic phase-field approach. Transmission electron microscopy reveals distinct bubble characteristics: the (Ni,Cr)O oxide/matrix interface exhibits the highest bubble density, while both the interface and the oxide interior share similar, fine bubble sizes (∼2.4 nm). In contrast, the Ni matrix hosts larger bubbles (∼3.8 nm) at a lower density. The anisotropic phase-field model, developed to quantitatively interpret these observations, identifies the coherent oxide interface as a potent defect sink. Its strong absorption of helium and vacancies promotes dense bubble nucleation while inducing a local vacancy depletion that suppresses coarsening in the oxide. Conversely, in the matrix, long-range diffusion enables the growth of sparse, large bubbles. Critically, the simulation reproduces the experimentally observed quadrilateral bubble morphology in the matrix, confirming the dominance of surface energy anisotropy in shaping bubbles. This work provides the first quantitative interpretation of anisotropic bubble evolution at oxide interfaces in IN617, highlighting the critical interplay between defect sink strength and crystallographic anisotropy. The coupled methodology establishes a predictive framework for designing radiation-tolerant microstructures.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156417"},"PeriodicalIF":3.2,"publicationDate":"2025-12-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882273","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 : 2025-12-27DOI: 10.1016/j.jnucmat.2025.156409
J.T. Rizk, X.-Y. Liu, D.A. Andersson, E. Kardoulaki, N.M. Abdul-Jabbar
The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. The DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.
{"title":"Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds","authors":"J.T. Rizk, X.-Y. Liu, D.A. Andersson, E. Kardoulaki, N.M. Abdul-Jabbar","doi":"10.1016/j.jnucmat.2025.156409","DOIUrl":"10.1016/j.jnucmat.2025.156409","url":null,"abstract":"<div><div>The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard <em>U</em> model (DFT+<em>U</em>) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+<em>U</em> calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. The DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156409"},"PeriodicalIF":3.2,"publicationDate":"2025-12-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922719","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 : 2025-12-26DOI: 10.1016/j.jnucmat.2025.156416
Calum S. Cunningham, Georgios Papanikos
Predicting irradiation-induced embrittlement of a nuclear reactor pressure vessel (RPV) is paramount to safe operation. Machine learning (ML) is an emerging statistical analysis tool with the potential to improve on conventional predictive models. Developing a supervised ML model requires many user-dependent decisions during data processing, not least the selection of a database, many of which differ amongst past studies. This work investigates a range of different procedures for developing an ML model to predict the transition temperature shift (ΔT41J), with the aim of identifying the best approaches to support the generation of future best practice guidelines. The impacts of different pre-processing and data selection approaches are evaluated, including comparison with conventional models and particular focus on the ability of ML models to extrapolate beyond the training data domain. A multi-layer perceptron ensemble model is found to be optimal and capable of producing more accurate predictions than a leading embrittlement trend curve. It is shown that differences in pre processing choices are not highly influential on final models. However, training data selection is critically important due to the unreliability of ML models when extrapolating. Recommendations are provided to maximise the benefit of ML in predicting RPV structural integrity and to support the adoption of safe, reliable best practice guidance in codes and standards.
{"title":"Using machine learning to predict reactor pressure vessel embrittlement: Human factors and best practice","authors":"Calum S. Cunningham, Georgios Papanikos","doi":"10.1016/j.jnucmat.2025.156416","DOIUrl":"10.1016/j.jnucmat.2025.156416","url":null,"abstract":"<div><div>Predicting irradiation-induced embrittlement of a nuclear reactor pressure vessel (RPV) is paramount to safe operation. Machine learning (ML) is an emerging statistical analysis tool with the potential to improve on conventional predictive models. Developing a supervised ML model requires many user-dependent decisions during data processing, not least the selection of a database, many of which differ amongst past studies. This work investigates a range of different procedures for developing an ML model to predict the transition temperature shift (ΔT<sub>41J</sub>), with the aim of identifying the best approaches to support the generation of future best practice guidelines. The impacts of different pre-processing and data selection approaches are evaluated, including comparison with conventional models and particular focus on the ability of ML models to extrapolate beyond the training data domain. A multi-layer perceptron ensemble model is found to be optimal and capable of producing more accurate predictions than a leading embrittlement trend curve. It is shown that differences in pre processing choices are not highly influential on final models. However, training data selection is critically important due to the unreliability of ML models when extrapolating. Recommendations are provided to maximise the benefit of ML in predicting RPV structural integrity and to support the adoption of safe, reliable best practice guidance in codes and standards.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156416"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145957738","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 : 2025-12-26DOI: 10.1016/j.jnucmat.2025.156410
Ine Arts , Rolando Saniz , Gianguido Baldinozzi , Gregory Leinders , Marc Verwerft , Dirk Lamoen
The oxidation of UO2 is primarily governed by the diffusion of oxygen through the lattice. Oxygen diffusion is significantly influenced by defects and interfaces, with grain boundaries being particularly relevant in spent nuclear fuel due to their increasing concentration at the periphery of fuel rods. While experimental studies on bulk uranium suggest defects enhance oxygen diffusion, the role of grain boundaries remains contentious, with discrepancies between theoretical predictions of enhanced diffusion and experimental observations. This study employs density functional theory (DFT+U) to investigate the electronic properties and stability of interstitial oxygen in two coincident site lattice grain boundaries, Σ3 {111} and Σ5 {210}, in UO2. We compare stoichiometric and non-stoichiometric grain boundary models, examining their formation energies, defect interactions, and local structural distortions. The interstitial oxygen defects cause an expansion of the oxygen cage and a contraction of the U-O bonds, both in bulk and at the grain boundaries. The Σ3 grain boundary showed potential for defect accumulation, while the Σ5 grain boundary did not demonstrate reduced defect formation energies relative to the bulk. Our findings contribute to the understanding the UO2 oxidation process, in an attempt to address inconsistencies between theoretical and experimental studies on oxygen diffusion in grain boundaries.
{"title":"Electronic properties and stability of interstitial oxygen in UO2 grain boundaries: An ab initio study","authors":"Ine Arts , Rolando Saniz , Gianguido Baldinozzi , Gregory Leinders , Marc Verwerft , Dirk Lamoen","doi":"10.1016/j.jnucmat.2025.156410","DOIUrl":"10.1016/j.jnucmat.2025.156410","url":null,"abstract":"<div><div>The oxidation of UO<sub>2</sub> is primarily governed by the diffusion of oxygen through the lattice. Oxygen diffusion is significantly influenced by defects and interfaces, with grain boundaries being particularly relevant in spent nuclear fuel due to their increasing concentration at the periphery of fuel rods. While experimental studies on bulk uranium suggest defects enhance oxygen diffusion, the role of grain boundaries remains contentious, with discrepancies between theoretical predictions of enhanced diffusion and experimental observations. This study employs density functional theory (DFT+U) to investigate the electronic properties and stability of interstitial oxygen in two coincident site lattice grain boundaries, Σ3 {111} and Σ5 {210}, in UO<sub>2</sub>. We compare stoichiometric and non-stoichiometric grain boundary models, examining their formation energies, defect interactions, and local structural distortions. The interstitial oxygen defects cause an expansion of the oxygen cage and a contraction of the U-O bonds, both in bulk and at the grain boundaries. The Σ3 grain boundary showed potential for defect accumulation, while the Σ5 grain boundary did not demonstrate reduced defect formation energies relative to the bulk. Our findings contribute to the understanding the UO<sub>2</sub> oxidation process, in an attempt to address inconsistencies between theoretical and experimental studies on oxygen diffusion in grain boundaries.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156410"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882270","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 : 2025-12-26DOI: 10.1016/j.jnucmat.2025.156414
Canjia Huang , Zifeng Deng , Xingli Wang , Qiang Li , Wanjing Wang , Zhilu Liu , Jieyao He , Jiaxin Jin , Wei Cao , Zongxiao Guo , Fan Wang , Yunming Qiu , Ying Liu , Chunyan Yu , Shixing Wang , Jianjun Huang
The fabrication of high-quality W/steel first wall of blanket poses significant challenges in the field of nuclear fusion, primarily due to the substantial differences in thermodynamic properties between tungsten (W) and stainless steel (SS). Detonation sprayed (DS) iron (Fe) coating is a good alternative thermal stress-relieving interlayer between the DS-W armor and the steel substrate. This study adopted two post-heat treatment strategies to optimize the microstructure and improve the performance of the DS-Fe interlayer: carrying out vacuum heat treatments (ranging from 500 °C to 900 °C) on the DS-W/DS-Fe/SS structure and the DS-Fe/SS structure following the deposition of DS-W coating, respectively. The microstructures of the DS-Fe coatings and the DS-W/DS-Fe/SS structures after the heat treatments showed that three sets of these heat-treated specimens had better microstructure improvement than others. The electron beam-based transient high heat flux (HHF) loading tests revealed that the heat-treated specimens demonstrated notable improvements in the performance of the coating structure. Moreover, under transient HHF, the specimens subjected to 900 °C heat treatment after the deposition of DS-W coating exhibited remarkable performance, followed by those treated at 900 °C before the deposition of DS-W coating. The differences in the damage behaviors of the coating structures at 50 MJ/m² HHF between the two sets were studied in order to understand their performance characteristics and distinctions. These results confirmed that holistic heat treatment on the DS-W/DS-Fe/SS structure effectively alleviated the thermal stress in it under the fusion reactor conditions, thereby providing critical references for the manufacturing of the blanket first wall.
{"title":"Influence of heat treatment on the microstructure and performance of detonation sprayed W/Fe/steel first wall structure","authors":"Canjia Huang , Zifeng Deng , Xingli Wang , Qiang Li , Wanjing Wang , Zhilu Liu , Jieyao He , Jiaxin Jin , Wei Cao , Zongxiao Guo , Fan Wang , Yunming Qiu , Ying Liu , Chunyan Yu , Shixing Wang , Jianjun Huang","doi":"10.1016/j.jnucmat.2025.156414","DOIUrl":"10.1016/j.jnucmat.2025.156414","url":null,"abstract":"<div><div>The fabrication of high-quality W/steel first wall of blanket poses significant challenges in the field of nuclear fusion, primarily due to the substantial differences in thermodynamic properties between tungsten (W) and stainless steel (SS). Detonation sprayed (DS) iron (Fe) coating is a good alternative thermal stress-relieving interlayer between the DS-W armor and the steel substrate. This study adopted two post-heat treatment strategies to optimize the microstructure and improve the performance of the DS-Fe interlayer: carrying out vacuum heat treatments (ranging from 500 °C to 900 °C) on the DS-W/DS-Fe/SS structure and the DS-Fe/SS structure following the deposition of DS-W coating, respectively. The microstructures of the DS-Fe coatings and the DS-W/DS-Fe/SS structures after the heat treatments showed that three sets of these heat-treated specimens had better microstructure improvement than others. The electron beam-based transient high heat flux (HHF) loading tests revealed that the heat-treated specimens demonstrated notable improvements in the performance of the coating structure. Moreover, under transient HHF, the specimens subjected to 900 °C heat treatment after the deposition of DS-W coating exhibited remarkable performance, followed by those treated at 900 °C before the deposition of DS-W coating. The differences in the damage behaviors of the coating structures at 50 MJ/m² HHF between the two sets were studied in order to understand their performance characteristics and distinctions. These results confirmed that holistic heat treatment on the DS-W/DS-Fe/SS structure effectively alleviated the thermal stress in it under the fusion reactor conditions, thereby providing critical references for the manufacturing of the blanket first wall.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156414"},"PeriodicalIF":3.2,"publicationDate":"2025-12-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882338","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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