Pub Date : 2026-01-15DOI: 10.1016/j.jnucmat.2026.156462
Hongcai Liang , Yi Wang , Wensheng Lai
A machine learning interatomic potential based on the moment tensor framework (moment tensor potential, MTP) for the uranium-molybdenum-zirconium ternary system is developed to investigate the plastic deformation mechanisms of U-Mo-Zr alloys. The MTP was trained using an extensive reference database generated from density functional theory calculations, and the resulting potential accurately reproduces the fundamental properties of the γ-phase UMo alloy. Molecular dynamics simulations of the Peierls-Nabarro stresses, modified by the generalized stacking fault energy, reveal that the {110}⟨111⟩ slip system is preferred over the {112}⟨111⟩ and {123}⟨111⟩ systems, and this preference weakens with increasing Zr content. Under uniaxial stretching along the [110] direction, the activation of the {110}⟨111⟩ slip system and the formation of corresponding slip bands are observed. Calculations of the generalized planar fault energy suggest that twinning is more favorable than slip in the {112}⟨111⟩ system, and such priority is enhanced with increasing Zr content. Accordingly, twinning rather than slip is predicted to occur during plastic deformation under certain conditions, which is confirmed by the activation of the {112}⟨111⟩ twinning system during uniaxial stretching along the [100] direction. These findings provide theoretical insights that may promote the processing and application of uranium-based metallic nuclear fuels in future advanced reactor systems.
{"title":"Elucidating plastic deformation mechanisms in γ-phase U-Mo-Zr ternary alloys using a machine-learning moment tensor potential","authors":"Hongcai Liang , Yi Wang , Wensheng Lai","doi":"10.1016/j.jnucmat.2026.156462","DOIUrl":"10.1016/j.jnucmat.2026.156462","url":null,"abstract":"<div><div>A machine learning interatomic potential based on the moment tensor framework (moment tensor potential, MTP) for the uranium-molybdenum-zirconium ternary system is developed to investigate the plastic deformation mechanisms of U-Mo-Zr alloys. The MTP was trained using an extensive reference database generated from density functional theory calculations, and the resulting potential accurately reproduces the fundamental properties of the <em>γ</em>-phase UMo alloy. Molecular dynamics simulations of the Peierls-Nabarro stresses, modified by the generalized stacking fault energy, reveal that the {110}⟨111⟩ slip system is preferred over the {112}⟨111⟩ and {123}⟨111⟩ systems, and this preference weakens with increasing Zr content. Under uniaxial stretching along the [110] direction, the activation of the {110}⟨111⟩ slip system and the formation of corresponding slip bands are observed. Calculations of the generalized planar fault energy suggest that twinning is more favorable than slip in the {112}⟨111⟩ system, and such priority is enhanced with increasing Zr content. Accordingly, twinning rather than slip is predicted to occur during plastic deformation under certain conditions, which is confirmed by the activation of the {112}⟨111⟩ twinning system during uniaxial stretching along the [100] direction. These findings provide theoretical insights that may promote the processing and application of uranium-based metallic nuclear fuels in future advanced reactor systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156462"},"PeriodicalIF":3.2,"publicationDate":"2026-01-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024750","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 : 2026-01-13DOI: 10.1016/j.jnucmat.2026.156460
Sara E. Gilson , Kathryn M. Lawson , Thomas Dyke , Kayron Rogers , Tamara Keever , Andrew Miskowiec , Tyler L. Spano
The availability of actinide standard materials for use in nuclear safeguard applications is critical, as is thorough characterization thereof. Although accurate trace element compositions and isotopic considerations are paramount for deployment of reference standards, structural characterization is also essential towards accurately describing the chemical form and potential matrix effects in candidate materials. To this end, samples of NpO2 were synthesized via a direct denitration (DD) method and probed with powder X-ray diffraction (PXRD), Raman spectroscopy, and scanning electron microscopy (SEM) for structural and morphological characterization and comparison with NpO2 materials produced via modified direct denitration (MDD). PXRD confirmed the bulk identity of NpO2, and no additional phases were identified using this method. Analysis of Raman data collected using a 532 nm excitation wavelength indicates that samples are mostly phase pure; however, some variability in spectral features is observed. Analysis of additional spectroscopic data collected with a 785 nm excitation wavelength revealed variability in the relative intensity of spectral features. Raman spectroscopy indicates that the sample is primarily NpO2; however, additional signals indicate possible structural disorder, oxidized species, or potential contributions from other Np phases. To further investigate the possibility of additional phase contributions within the sample of NpO2, Raman spectroscopic mapping was employed to examine the homogeneity of the sample produced via DD. From this analysis, we determined that despite variability in the intensity of Raman-active vibrational modes, consistent spectra are obtained throughout the area of the sample investigated. SEM images show aggregates with variable sizes and shapes, with rounded, primary particles possessing an average diameter of approximately 100 nm. Comparison of the results of these multimodal analyses to the literature indicates that the crystal chemical, spectroscopic, and microstructural properties of NpO2 vary based on synthesis method, even if X-ray diffraction data indicate that the bulk phase is NpO2.
{"title":"Understanding the structural and morphological effects of synthesis route on NpO2","authors":"Sara E. Gilson , Kathryn M. Lawson , Thomas Dyke , Kayron Rogers , Tamara Keever , Andrew Miskowiec , Tyler L. Spano","doi":"10.1016/j.jnucmat.2026.156460","DOIUrl":"10.1016/j.jnucmat.2026.156460","url":null,"abstract":"<div><div>The availability of actinide standard materials for use in nuclear safeguard applications is critical, as is thorough characterization thereof. Although accurate trace element compositions and isotopic considerations are paramount for deployment of reference standards, structural characterization is also essential towards accurately describing the chemical form and potential matrix effects in candidate materials. To this end, samples of NpO<sub>2</sub> were synthesized via a direct denitration (DD) method and probed with powder X-ray diffraction (PXRD), Raman spectroscopy, and scanning electron microscopy (SEM) for structural and morphological characterization and comparison with NpO<sub>2</sub> materials produced via modified direct denitration (MDD). PXRD confirmed the bulk identity of NpO<sub>2</sub>, and no additional phases were identified using this method. Analysis of Raman data collected using a 532 nm excitation wavelength indicates that samples are mostly phase pure; however, some variability in spectral features is observed. Analysis of additional spectroscopic data collected with a 785 nm excitation wavelength revealed variability in the relative intensity of spectral features. Raman spectroscopy indicates that the sample is primarily NpO<sub>2</sub>; however, additional signals indicate possible structural disorder, oxidized species, or potential contributions from other Np phases. To further investigate the possibility of additional phase contributions within the sample of NpO<sub>2</sub>, Raman spectroscopic mapping was employed to examine the homogeneity of the sample produced via DD. From this analysis, we determined that despite variability in the intensity of Raman-active vibrational modes, consistent spectra are obtained throughout the area of the sample investigated. SEM images show aggregates with variable sizes and shapes, with rounded, primary particles possessing an average diameter of approximately 100 nm. Comparison of the results of these multimodal analyses to the literature indicates that the crystal chemical, spectroscopic, and microstructural properties of NpO<sub>2</sub> vary based on synthesis method, even if X-ray diffraction data indicate that the bulk phase is NpO<sub>2</sub>.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156460"},"PeriodicalIF":3.2,"publicationDate":"2026-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024667","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 : 2026-01-13DOI: 10.1016/j.jnucmat.2026.156459
Wenliang Xu, Shushan Cui, Shilv Yu, Mengsheng Zhai, Sheng Zhang, Denglei Chen, Yun Fang, Pei Zhang, Dawu Xiao, Tao Fa
Tensile specimens of lamellar U-2Nb alloy were prepared using three different cutting methods: fast-speed wire electrical discharge machining (FS-WEDM), slow-speed wire electrical discharge machining (SS-WEDM) and milling. Significant discrepancies in mechanical performance were observed, with the FS-WEDM and SS-WEDM specimens exhibiting total elongations below 10%, which is substantially lower than that achieved by milling. Thermal desorption analysis and fractographic examination indicated that hydrogen trapped at α-U (or UH3) phases and α/γ1-2 interfaces is responsible for the hydrogen embrittlement (HE) observed in the WEDM-processed specimens. This conclusion was confirmed by subsequent vacuum dehydrogenation treatments at 393 K, 573 K and 873 K, coupled with corresponding fractographic analysis. Consequently, the utilizing of WEDM should be extremely cautious when preparing tensile specimens for U-2Nb alloy, and milling is recommended as the preferred alternative. This study reveals that WEDM introduces hydrogen, leading to embrittlement of U-2Nb alloy, and provides critical guidance for the sample preparation of HE-sensitive materials.
{"title":"Processing dictates properties: Mitigating hydrogen embrittlement in U-2Nb alloy by cutting method selection","authors":"Wenliang Xu, Shushan Cui, Shilv Yu, Mengsheng Zhai, Sheng Zhang, Denglei Chen, Yun Fang, Pei Zhang, Dawu Xiao, Tao Fa","doi":"10.1016/j.jnucmat.2026.156459","DOIUrl":"10.1016/j.jnucmat.2026.156459","url":null,"abstract":"<div><div>Tensile specimens of lamellar U-2Nb alloy were prepared using three different cutting methods: fast-speed wire electrical discharge machining (FS-WEDM), slow-speed wire electrical discharge machining (SS-WEDM) and milling. Significant discrepancies in mechanical performance were observed, with the FS-WEDM and SS-WEDM specimens exhibiting total elongations below 10%, which is substantially lower than that achieved by milling. Thermal desorption analysis and fractographic examination indicated that hydrogen trapped at α-U (or UH<sub>3</sub>) phases and α/γ<sub>1-2</sub> interfaces is responsible for the hydrogen embrittlement (HE) observed in the WEDM-processed specimens. This conclusion was confirmed by subsequent vacuum dehydrogenation treatments at 393 K, 573 K and 873 K, coupled with corresponding fractographic analysis. Consequently, the utilizing of WEDM should be extremely cautious when preparing tensile specimens for U-2Nb alloy, and milling is recommended as the preferred alternative. This study reveals that WEDM introduces hydrogen, leading to embrittlement of U-2Nb alloy, and provides critical guidance for the sample preparation of HE-sensitive materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156459"},"PeriodicalIF":3.2,"publicationDate":"2026-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024749","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 : 2026-01-13DOI: 10.1016/j.jnucmat.2026.156461
Jiyong Huang , Hucheng Yu , Yifan Ding , Xiangbing Liu , Ziqi Cao , Runzhong Wang , Wenqing Jia , Sheng Fan , Guang Ran
Understanding the formation and evolution of irradiation-induced defects under different effective sink strengths is essential for designing radiation-resistant structural materials in advanced nuclear systems. In this study, in situ He ion irradiation experiments were conducted at 450 °C on two regions of low-alloy steel materials with distinct sink strengths characterized by variations in dislocation density and precipitate distribution. In situ TEM observations revealed the formation and evolution behaviors of dislocation loops and helium bubbles during irradiation and subsequent annealing at 700 and 750 °C. Regions with higher sink strength exhibited lower dislocation loop density, smaller loop size, and a reduced fraction of a < 100> loops, indicating enhanced point-defect recombination and suppressed loop nucleation and growth. During irradiation, helium bubbles in the low-sink region were larger and non-spherical, while those in the high-sink region were finer and denser. Upon annealing, bubble size reversal occurred: bubbles in the high-sink region grew rapidly at later stages. At the final annealing stage, bubbles in the high-sink region were largely pinned, whereas smaller bubbles in the low-sink region exhibited thermally driven migration. These findings elucidate the critical role of effective sink strength in defect evolution and provide experimental insights for the microstructural design of radiation-resistant materials.
{"title":"Influence of defect sink strength on dislocation loop and helium bubble evolution in low-alloy steel under in-situ He irradiation and annealing","authors":"Jiyong Huang , Hucheng Yu , Yifan Ding , Xiangbing Liu , Ziqi Cao , Runzhong Wang , Wenqing Jia , Sheng Fan , Guang Ran","doi":"10.1016/j.jnucmat.2026.156461","DOIUrl":"10.1016/j.jnucmat.2026.156461","url":null,"abstract":"<div><div>Understanding the formation and evolution of irradiation-induced defects under different effective sink strengths is essential for designing radiation-resistant structural materials in advanced nuclear systems. In this study, in situ He ion irradiation experiments were conducted at 450 °C on two regions of low-alloy steel materials with distinct sink strengths characterized by variations in dislocation density and precipitate distribution. In situ TEM observations revealed the formation and evolution behaviors of dislocation loops and helium bubbles during irradiation and subsequent annealing at 700 and 750 °C. Regions with higher sink strength exhibited lower dislocation loop density, smaller loop size, and a reduced fraction of <em>a</em> < 100> loops, indicating enhanced point-defect recombination and suppressed loop nucleation and growth. During irradiation, helium bubbles in the low-sink region were larger and non-spherical, while those in the high-sink region were finer and denser. Upon annealing, bubble size reversal occurred: bubbles in the high-sink region grew rapidly at later stages. At the final annealing stage, bubbles in the high-sink region were largely pinned, whereas smaller bubbles in the low-sink region exhibited thermally driven migration. These findings elucidate the critical role of effective sink strength in defect evolution and provide experimental insights for the microstructural design of radiation-resistant materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156461"},"PeriodicalIF":3.2,"publicationDate":"2026-01-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976098","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 : 2026-01-12DOI: 10.1016/j.jnucmat.2026.156450
Omeet N. Patel , Dwaipayan Dasgupta , Asanka Weerasinghe , Brian D. Wirth , Dimitrios Maroudas
Recent experiments have revealed a strong dependence of surface morphology on surface crystallographic orientation during the initial phase of ‘fuzz’ growth in fusion plasma-facing tungsten. Here, we examine this dependence using predictions of dynamical simulations based on an atomistically-informed continuum-scale model of surface evolution in plasma-irradiated tungsten (W). Upon exposure to a low-energy helium plasma, tungsten surface adatoms are produced as a result of surface vacancy-adatom pair formation and the flux of tungsten self-interstitial atoms, resulting from growing over-pressurized helium bubbles, toward the plasma-exposed surface. A combination of a stress-governed surface morphological instability, the preferential diffusion of tungsten surface adatoms, and the surface free energy anisotropy leads to the growth of various features on the plasma-facing surface. Simulations of the morphological response of helium irradiated W(111), W(110), and W(100) surfaces show the growth of triangular, stripe-shaped, and pyramidal surface features, respectively, and predict the growth kinetics, namely, the thickness evolution of a damaged surface layer at the early stage of fuzz formation, consistent with experimental observations. This work elucidates the effect of surface crystallographic orientation on the morphological evolution of plasma-facing tungsten surfaces, where the stress-governed surface morphological instability drives the formation of surface features (stripes or mounds), the preferential surface adatom migration controls the anisotropic growth of these features, while the surface free energy anisotropy is responsible for the formation of planar facets on the features emanating from the plasma-facing surface. We observe that the average separation between the resulting surface features increases with increasing temperature and duration of plasma exposure. This coarsening process is thermally activated, with the temperature dependence of the average feature separation well described by an Arrhenius relationship. Our modeling framework enables a predictive assessment of fuzz onset in plasma-facing tungsten across different crystallographic orientations.
{"title":"Surface morphological response of plasma-facing tungsten: Effects of surface crystallographic orientation and prediction of fuzz onset","authors":"Omeet N. Patel , Dwaipayan Dasgupta , Asanka Weerasinghe , Brian D. Wirth , Dimitrios Maroudas","doi":"10.1016/j.jnucmat.2026.156450","DOIUrl":"10.1016/j.jnucmat.2026.156450","url":null,"abstract":"<div><div>Recent experiments have revealed a strong dependence of surface morphology on surface crystallographic orientation during the initial phase of ‘fuzz’ growth in fusion plasma-facing tungsten. Here, we examine this dependence using predictions of dynamical simulations based on an atomistically-informed continuum-scale model of surface evolution in plasma-irradiated tungsten (W). Upon exposure to a low-energy helium plasma, tungsten surface adatoms are produced as a result of surface vacancy-adatom pair formation and the flux of tungsten self-interstitial atoms, resulting from growing over-pressurized helium bubbles, toward the plasma-exposed surface. A combination of a stress-governed surface morphological instability, the preferential diffusion of tungsten surface adatoms, and the surface free energy anisotropy leads to the growth of various features on the plasma-facing surface. Simulations of the morphological response of helium irradiated W(111), W(110), and W(100) surfaces show the growth of triangular, stripe-shaped, and pyramidal surface features, respectively, and predict the growth kinetics, namely, the thickness evolution of a damaged surface layer at the early stage of fuzz formation, consistent with experimental observations. This work elucidates the effect of surface crystallographic orientation on the morphological evolution of plasma-facing tungsten surfaces, where the stress-governed surface morphological instability drives the formation of surface features (stripes or mounds), the preferential surface adatom migration controls the anisotropic growth of these features, while the surface free energy anisotropy is responsible for the formation of planar facets on the features emanating from the plasma-facing surface. We observe that the average separation between the resulting surface features increases with increasing temperature and duration of plasma exposure. This coarsening process is thermally activated, with the temperature dependence of the average feature separation well described by an Arrhenius relationship. Our modeling framework enables a predictive assessment of fuzz onset in plasma-facing tungsten across different crystallographic orientations.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156450"},"PeriodicalIF":3.2,"publicationDate":"2026-01-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976102","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 : 2026-01-10DOI: 10.1016/j.jnucmat.2026.156451
Zijian Lin, Changjiang Hu, Yifan Li, Sinan Chen, Minzhang Lin, Jun Ma
The predication of advanced alloy corrosion risk by the H2O2 generation during coolant radiolysis has been an important issue in Pressurized water reactor (PWR) maintenance but remains rarely explored. Herein, this study developed the radiolysis and mixed-potential modelling (MPM) based on 60Co γ-irradiation products analysis together with electrochemical corrosion experiments conducted under non-irradiated conditions. First, the experimental data correlates the H2O2 production with a wide array of influencing practical factors such as temperature, LiOH/NH3/H3BO3 concentrations, providing a radiolysis model for corrosion basis. We next conducted the electrochemical corrosion behavior of FeCrAl (APMT and C26M) and Zircaloy–4 (Zr–4) alloys in simulated PWR coolants to develop MPM. The modeling predicts that hydrogen injections can reduce the Electrochemical Corrosion Potential (ECP). Under identical condition, Zr–4 exhibits the most positive ECP, while APMT shows the most negative value. Threshold analysis suggests that injecting approximately 5 mL (STP)/kg H2 can lower the ECP of all tested alloys below −0.23 V (vs. SHE) that can prevent stress corrosion cracking risk.
利用冷却剂辐射分解过程中产生的H2O2来预测先进合金的腐蚀风险一直是压水堆（PWR）维护中的一个重要问题，但很少有人对此进行研究。在此基础上，本研究基于60Co γ辐照产物分析和非辐照条件下的电化学腐蚀实验，建立了辐射分解和混合电位模型（MPM）。首先，实验数据将H2O2产量与温度、LiOH/NH3/H3BO3浓度等一系列影响实际因素联系起来，为腐蚀基础提供了辐射分解模型。接下来，我们对FeCrAl （APMT和C26M）和Zircaloy-4 （Zr-4）合金在模拟压水堆冷却剂中的电化学腐蚀行为进行了研究。模型预测，氢注入可以降低电化学腐蚀电位（ECP）。在相同条件下，Zr-4的ECP为正，APMT的ECP为负。阈值分析表明，注入约5 mL (STP)/kg H2可以将所有测试合金的ECP降低到- 0.23 V以下（相对于SHE），从而可以防止应力腐蚀开裂风险。
{"title":"Prediction of electrochemical corrosion potential in PWR coolant for advanced FeCrAl and Zr–4 alloys via radiolysis and mixed potential modelling","authors":"Zijian Lin, Changjiang Hu, Yifan Li, Sinan Chen, Minzhang Lin, Jun Ma","doi":"10.1016/j.jnucmat.2026.156451","DOIUrl":"10.1016/j.jnucmat.2026.156451","url":null,"abstract":"<div><div>The predication of advanced alloy corrosion risk by the H<sub>2</sub>O<sub>2</sub> generation during coolant radiolysis has been an important issue in Pressurized water reactor (PWR) maintenance but remains rarely explored. Herein, this study developed the radiolysis and mixed-potential modelling (MPM) based on <sup>60</sup>Co γ-irradiation products analysis together with electrochemical corrosion experiments conducted under non-irradiated conditions. First, the experimental data correlates the H<sub>2</sub>O<sub>2</sub> production with a wide array of influencing practical factors such as temperature, LiOH/NH<sub>3</sub>/H<sub>3</sub>BO<sub>3</sub> concentrations, providing a radiolysis model for corrosion basis. We next conducted the electrochemical corrosion behavior of FeCrAl (APMT and C26M) and Zircaloy–4 (Zr–4) alloys in simulated PWR coolants to develop MPM. The modeling predicts that hydrogen injections can reduce the Electrochemical Corrosion Potential (ECP). Under identical condition, Zr–4 exhibits the most positive ECP, while APMT shows the most negative value. Threshold analysis suggests that injecting approximately 5 mL (STP)/kg H<sub>2</sub> can lower the ECP of all tested alloys below −0.23 V (vs. SHE) that can prevent stress corrosion cracking risk.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156451"},"PeriodicalIF":3.2,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976101","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 : 2026-01-09DOI: 10.1016/j.jnucmat.2026.156443
Shuang Hu , Zhen Wu , Yao Yu , Mei Zhou , Qigui Yang , Peng Zhang , Yu Chen , Mingpan Wan , Te Zhu , Xingzhong Cao
Although titanium alloys have gained significant attention for their potential applications in advanced reactors, experimental studies on irradiation damage under varied irradiation conditions remain insufficient, limiting the understanding of defect evolution and hardening behavior. This study selected the near α titanium alloy Ti-5Al-3V-3Zr-Cr (Ti-5331), which has ideal mechanical properties, and compares its irradiation-induced defect formation, softening, and hardening effects under different fluences and temperatures. Results from slow positron-beam Doppler broadening spectroscopy (DBS) confirm that hydrogen ion irradiation generates a significant number of vacancy-type defects and HmVn complexes in room temperature (RT) and high temperatures (473 K and 573 K). At a high fluence (RT-1 × 1017 H⁺/cm²), the excess HmVn complexes will inhibit the increase in the S parameter. In contrast, at 473 K and 573 K, thermal activation reduces the concentration of vacancy-type defects, and led to a significant decrease in the overall S parameter. In addition to the aforementioned defects, a large number of hydrogen atoms occupying vacancies gradually form small hydrogen bubbles, which increase in size with increasing fluence (5 × 1016 H⁺/cm² to 1 × 1017 H⁺/cm²) and temperature (RT to 573 K). Notably, the hydrogen bubbles in the α phase are larger than those in the β phase (e.g. RT-1 × 1017 H⁺/cm² sample). Unlike the typical irradiation hardening phenomenon, the nanoindentation results exhibit significant irradiation softening. The softening effect becomes more pronounced with increasing room-temperature irradiation fluence, resulting in a hardness reduction of up to 19% compared to the unirradiated samples. Irradiation at elevated temperatures also resulted in significant softening. The softening effect may be attributed to hydrogen-induced local plastic deformation, where hydrogen enhances the interaction of dislocations on different slip planes, leading to the increased complexity of dislocation structures and increased local plasticity. These findings elucidate hydrogen-defect interactions and temperature-fluence synergies, critical for designing irradiation-resistant titanium alloys in nuclear applications.
{"title":"Hydrogen ion irradiation-induced defect evolution and softening in near-α Ti-5331 alloy: Effects of fluence and temperature","authors":"Shuang Hu , Zhen Wu , Yao Yu , Mei Zhou , Qigui Yang , Peng Zhang , Yu Chen , Mingpan Wan , Te Zhu , Xingzhong Cao","doi":"10.1016/j.jnucmat.2026.156443","DOIUrl":"10.1016/j.jnucmat.2026.156443","url":null,"abstract":"<div><div>Although titanium alloys have gained significant attention for their potential applications in advanced reactors, experimental studies on irradiation damage under varied irradiation conditions remain insufficient, limiting the understanding of defect evolution and hardening behavior. This study selected the near α titanium alloy Ti-5Al-3V-3Zr-Cr (Ti-5331), which has ideal mechanical properties, and compares its irradiation-induced defect formation, softening, and hardening effects under different fluences and temperatures. Results from slow positron-beam Doppler broadening spectroscopy (DBS) confirm that hydrogen ion irradiation generates a significant number of vacancy-type defects and H<sub>m</sub>V<sub>n</sub> complexes in room temperature (RT) and high temperatures (473 K and 573 K). At a high fluence (RT-1 × 10<sup>17</sup> H⁺/cm²), the excess H<sub>m</sub>V<sub>n</sub> complexes will inhibit the increase in the S parameter. In contrast, at 473 K and 573 K, thermal activation reduces the concentration of vacancy-type defects, and led to a significant decrease in the overall S parameter. In addition to the aforementioned defects, a large number of hydrogen atoms occupying vacancies gradually form small hydrogen bubbles, which increase in size with increasing fluence (5 × 10<sup>16</sup> H⁺/cm² to 1 × 10<sup>17</sup> H⁺/cm²) and temperature (RT to 573 K). Notably, the hydrogen bubbles in the α phase are larger than those in the β phase (e.g. RT-1 × 10<sup>17</sup> H⁺/cm² sample). Unlike the typical irradiation hardening phenomenon, the nanoindentation results exhibit significant irradiation softening. The softening effect becomes more pronounced with increasing room-temperature irradiation fluence, resulting in a hardness reduction of up to 19% compared to the unirradiated samples. Irradiation at elevated temperatures also resulted in significant softening. The softening effect may be attributed to hydrogen-induced local plastic deformation, where hydrogen enhances the interaction of dislocations on different slip planes, leading to the increased complexity of dislocation structures and increased local plasticity. These findings elucidate hydrogen-defect interactions and temperature-fluence synergies, critical for designing irradiation-resistant titanium alloys in nuclear applications.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156443"},"PeriodicalIF":3.2,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024745","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 : 2026-01-09DOI: 10.1016/j.jnucmat.2026.156447
Congcong Zhao , Junhong Jia , Kui Wang , Xiaoyu Jiao , Runze Wei , Jian Wang , Jie Yang , Yaoqiang Tang , Qingqing Zhang , Zongyu Zhang
This study systematically investigates the critical influence of Fe³⁺ impurity ions on the corrosion behavior of boron carbide (B4C) -reinforced silicon carbide (SiC) composite ceramics in FLiNaK molten salt. SiC-B₄C composites were fabricated via hot-press sintering, and their corrosion characteristics were compared in both original and Fe³⁺-contaminated environments. The results indicate that in pure molten salt, the addition of B₄C promotes the formation of a dense, graphitized carbon layer on the surface, which significantly enhances corrosion resistance. In contrast, the strong oxidizing power of Fe³⁺ disrupts the formation and stability of this protective layer. This disruption results in a loose, porous surface structure, a substantial increase in corrosion layer thickness and the creation of pathways for the corrosive melt penetration. Thermodynamic analysis further confirms that FeF₃, as a potent oxidizing agent, preferentially reacts with B₄C. Moreover, Fe³⁺ is regenerated through a cyclic disproportionation reaction (3Fe²⁺→2Fe³⁺+Fe), which intensifies the corrosion process. This work provides critical theoretical support for material selection and service-life assessment of structural components in molten salt reactors under practical operating conditions.
{"title":"Effect and mechanism of Fe³⁺ impurities on the corrosion behavior of SiC-B₄C composites in molten FLiNaK salt","authors":"Congcong Zhao , Junhong Jia , Kui Wang , Xiaoyu Jiao , Runze Wei , Jian Wang , Jie Yang , Yaoqiang Tang , Qingqing Zhang , Zongyu Zhang","doi":"10.1016/j.jnucmat.2026.156447","DOIUrl":"10.1016/j.jnucmat.2026.156447","url":null,"abstract":"<div><div>This study systematically investigates the critical influence of Fe³⁺ impurity ions on the corrosion behavior of boron carbide (B<sub>4</sub>C) -reinforced silicon carbide (SiC) composite ceramics in FLiNaK molten salt. SiC-B₄C composites were fabricated via hot-press sintering, and their corrosion characteristics were compared in both original and Fe³⁺-contaminated environments. The results indicate that in pure molten salt, the addition of B₄C promotes the formation of a dense, graphitized carbon layer on the surface, which significantly enhances corrosion resistance. In contrast, the strong oxidizing power of Fe³⁺ disrupts the formation and stability of this protective layer. This disruption results in a loose, porous surface structure, a substantial increase in corrosion layer thickness and the creation of pathways for the corrosive melt penetration. Thermodynamic analysis further confirms that FeF₃, as a potent oxidizing agent, preferentially reacts with B₄C. Moreover, Fe³⁺ is regenerated through a cyclic disproportionation reaction (3Fe²⁺→2Fe³⁺+Fe), which intensifies the corrosion process. This work provides critical theoretical support for material selection and service-life assessment of structural components in molten salt reactors under practical operating conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156447"},"PeriodicalIF":3.2,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145957740","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 : 2026-01-09DOI: 10.1016/j.jnucmat.2026.156448
Yifan Liu , Ke Xu , Zhiye Tang , Chenxin Yan , Shuo Jin , Linyun Liang , Hong-Bo Zhou , Guang-Hong Lu
Interstitial hydrogen (H) clusters with rock salt structure exhibit energy stability in tungsten (W) and play a crucial role in enhancing its hardness. However, the underlying physical mechanisms and the specific hardening behavior remain unclear. To this end, we systematically investigate the effect of the H cluster on the slip behavior of an 1/2 [111] () edge dislocation in W by using the molecular dynamics method. We first study the slip of the edge dislocation in the absence of the H cluster, which reveals typical phonon drag control characteristics. Based on this, a slip drag coefficient B(T) is obtained, enabling accurate prediction of the dislocation mobility under various temperatures and stresses. In the presence of H clusters, the hardening effect in W is significantly enhanced. Notably, the geometric parameters of the H cluster, i.e., height and diameter, exert significant regulatory influence on the slip behavior of the edge dislocation through a quantitative correlation. Furthermore, the critical resolved shear stress (CRSS) displays a slight dependence on temperature within the range of 100 K-800 K, indicating that the dislocation motion is primarily governed by the geometry of the H cluster. These results provide new insights into the mechanisms of H-induced irradiation hardening in W, offering valuable data to support the development of high-performance W-based materials with enhanced irradiation resistance and long-term service stability.
{"title":"Molecular dynamics study on the effect of interstitial hydrogen clusters on the slip of edge dislocation in tungsten","authors":"Yifan Liu , Ke Xu , Zhiye Tang , Chenxin Yan , Shuo Jin , Linyun Liang , Hong-Bo Zhou , Guang-Hong Lu","doi":"10.1016/j.jnucmat.2026.156448","DOIUrl":"10.1016/j.jnucmat.2026.156448","url":null,"abstract":"<div><div>Interstitial hydrogen (H) clusters with rock salt structure exhibit energy stability in tungsten (W) and play a crucial role in enhancing its hardness. However, the underlying physical mechanisms and the specific hardening behavior remain unclear. To this end, we systematically investigate the effect of the H cluster on the slip behavior of an 1/2 [111] (<span><math><mrow><mn>1</mn><mover><mn>1</mn><mo>¯</mo></mover><mn>0</mn></mrow></math></span>) edge dislocation in W by using the molecular dynamics method. We first study the slip of the edge dislocation in the absence of the H cluster, which reveals typical phonon drag control characteristics. Based on this, a slip drag coefficient <em>B</em>(<em>T</em>) is obtained, enabling accurate prediction of the dislocation mobility under various temperatures and stresses. In the presence of H clusters, the hardening effect in W is significantly enhanced. Notably, the geometric parameters of the H cluster, i.e., height and diameter, exert significant regulatory influence on the slip behavior of the edge dislocation through a quantitative correlation. Furthermore, the critical resolved shear stress (CRSS) displays a slight dependence on temperature within the range of 100 K-800 K, indicating that the dislocation motion is primarily governed by the geometry of the H cluster. These results provide new insights into the mechanisms of H-induced irradiation hardening in W, offering valuable data to support the development of high-performance W-based materials with enhanced irradiation resistance and long-term service stability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156448"},"PeriodicalIF":3.2,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145976097","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 : 2026-01-08DOI: 10.1016/j.jnucmat.2026.156445
Mikhail A. Sokolov , Roger E. Stoller
In order to examine the attenuation of radiation damage through the thickness of an irradiated reactor pressure vessel (RPV), four segments were acquired from the Zion Unit 1 power plant RPV after the plant was decommissioned. The Zion Unit 1 RPV Beltline Weld Segment 1 was cut into seven blocks, consisting of five base metal and two beltline welds from the high fluence region of the segment. Through-wall test specimens were machined and tested. Specimens included those used for Charpy impact, Master Curve fracture toughness testing, and chemical analysis. The observed through-thickness ductile-to-brittle transition temperatures in the beltline weld deviated significantly from the expected behavior based on the attenuation of fast fluence as a function of depth into the RPV. Beginning at the inside surface, the 41-J Charpy transition temperature was either flat or slightly increasing until the ¾ -T location. The results of a simple, model-based analysis of the Zion beltline weld material that included the irradiation conditions and material chemistry were generally consistent with industry trend curves and the standard attenuation model, rather than the observed data. Although there was no archive material from the RPV available to permit measurement of the unirradiated properties, fracture toughness specimens fabricated from archive surveillance weld were used to obtain an estimate of the initial through-thickness values of the Charpy transition temperature. The Charpy shifts obtained using this approach were similarly in disagreement with the predictions of the US NRC Regulatory Guide 1.99, Rev. 2. However, testing of irradiated Charpy specimens taken from the RPV following post-irradiation annealing (10 hr. at 500 °C) provided a quite different estimate of the unirradiated properties which improved the agreement between the inferred through-thickness Charpy shifts and exponential attenuation model included in Regulatory Guide 1.99/2. The analysis of the Zion data and data obtained in previous post-mortem examinations of decommissioned RPVs indicates that more work is needed to understand the through-thickness properties of RPV materials in order to properly assess through-wall damage attenuation.
{"title":"Analysis of attenuation data from the decommissioned ZIon unit 1 reactor pressure vessel beltline weld","authors":"Mikhail A. Sokolov , Roger E. Stoller","doi":"10.1016/j.jnucmat.2026.156445","DOIUrl":"10.1016/j.jnucmat.2026.156445","url":null,"abstract":"<div><div>In order to examine the attenuation of radiation damage through the thickness of an irradiated reactor pressure vessel (RPV), four segments were acquired from the Zion Unit 1 power plant RPV after the plant was decommissioned. The Zion Unit 1 RPV Beltline Weld Segment 1 was cut into seven blocks, consisting of five base metal and two beltline welds from the high fluence region of the segment. Through-wall test specimens were machined and tested. Specimens included those used for Charpy impact, Master Curve fracture toughness testing, and chemical analysis. The observed through-thickness ductile-to-brittle transition temperatures in the beltline weld deviated significantly from the expected behavior based on the attenuation of fast fluence as a function of depth into the RPV. Beginning at the inside surface, the 41-J Charpy transition temperature was either flat or slightly increasing until the ¾ -T location. The results of a simple, model-based analysis of the Zion beltline weld material that included the irradiation conditions and material chemistry were generally consistent with industry trend curves and the standard attenuation model, rather than the observed data. Although there was no archive material from the RPV available to permit measurement of the unirradiated properties, fracture toughness specimens fabricated from archive surveillance weld were used to obtain an estimate of the initial through-thickness values of the Charpy transition temperature. The Charpy shifts obtained using this approach were similarly in disagreement with the predictions of the US NRC Regulatory Guide 1.99, Rev. 2. However, testing of irradiated Charpy specimens taken from the RPV following post-irradiation annealing (10 hr. at 500 °C) provided a quite different estimate of the unirradiated properties which improved the agreement between the inferred through-thickness Charpy shifts and exponential attenuation model included in Regulatory Guide 1.99/2. The analysis of the Zion data and data obtained in previous post-mortem examinations of decommissioned RPVs indicates that more work is needed to understand the through-thickness properties of RPV materials in order to properly assess through-wall damage attenuation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"623 ","pages":"Article 156445"},"PeriodicalIF":3.2,"publicationDate":"2026-01-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146024753","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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