Journal of Nuclear Materials最新文献_第4页

Effect of the transmutation element rhenium on the retention, desorption, and diffusion behaviors of hydrogen isotopes in tungsten

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

Journal of Nuclear Materials

Pub Date : 2025-02-08 DOI: 10.1016/j.jnucmat.2025.155693

Baolong Fan , Fei Sun , Qihang Liu , Xiao-Chun Li , Jipeng Zhu , Kang-Yan Lu , Hui-Long Yang , Hai-Shan Zhou , Rui Ding , Lai-Ma Luo , Yasuhisa Oya , Yucheng Wu

The 14.1 MeV neutron irradiation in fusion reactors can cause significant damage to tungsten (W), leading to defect formation and the creation of various transmutation elements, with rhenium (Re) being a principal product. Hydrogen isotopes are essential as fuel for fusion reactors, but the presence of transmuted Re complicates the behavior of hydrogen isotopes in W. This study systematically investigated the behavior of hydrogen isotopes in W and W-5%Re by molecular dynamics methods. The results indicate that Re in the W bulk significantly influences the retention, desorption, and diffusion behaviors of deuterium (D). Specifically, the inhibitory effect of Re on D retention decreases with increasing temperature, and the differences in D retention between W and W-5%Re are closely linked to vacancy concentrations. The addition of Re effectively reduces these vacancy concentrations, with even small amounts proving significantly in decreasing D retention. Furthermore, the amount of D desorbed from W is higher than that from W-5%Re at a desorption temperature of 500 K. The presence of Re also increases the diffusion coefficient of D in W, although this coefficient remains relatively stable regardless of Re concentration. Regarding depth distribution, D enrichment is more pronounced near the surface of W than in W-5%Re, closely linked to vacancy concentrations. As temperature increases, more D atoms diffuse into the material's interior, with Re in the bulk of W promoting this diffusion and mitigating surface damage. This study enhances understanding of the impact of Re on hydrogen isotope behavior and supports future assessments of tritium behavior under neutron irradiation conditions.

{"title":"Effect of the transmutation element rhenium on the retention, desorption, and diffusion behaviors of hydrogen isotopes in tungsten","authors":"Baolong Fan , Fei Sun , Qihang Liu , Xiao-Chun Li , Jipeng Zhu , Kang-Yan Lu , Hui-Long Yang , Hai-Shan Zhou , Rui Ding , Lai-Ma Luo , Yasuhisa Oya , Yucheng Wu","doi":"10.1016/j.jnucmat.2025.155693","DOIUrl":"10.1016/j.jnucmat.2025.155693","url":null,"abstract":"<div><div>The 14.1 MeV neutron irradiation in fusion reactors can cause significant damage to tungsten (W), leading to defect formation and the creation of various transmutation elements, with rhenium (Re) being a principal product. Hydrogen isotopes are essential as fuel for fusion reactors, but the presence of transmuted Re complicates the behavior of hydrogen isotopes in W. This study systematically investigated the behavior of hydrogen isotopes in W and W-5%Re by molecular dynamics methods. The results indicate that Re in the W bulk significantly influences the retention, desorption, and diffusion behaviors of deuterium (D). Specifically, the inhibitory effect of Re on D retention decreases with increasing temperature, and the differences in D retention between W and W-5%Re are closely linked to vacancy concentrations. The addition of Re effectively reduces these vacancy concentrations, with even small amounts proving significantly in decreasing D retention. Furthermore, the amount of D desorbed from W is higher than that from W-5%Re at a desorption temperature of 500 K. The presence of Re also increases the diffusion coefficient of D in W, although this coefficient remains relatively stable regardless of Re concentration. Regarding depth distribution, D enrichment is more pronounced near the surface of W than in W-5%Re, closely linked to vacancy concentrations. As temperature increases, more D atoms diffuse into the material's interior, with Re in the bulk of W promoting this diffusion and mitigating surface damage. This study enhances understanding of the impact of Re on hydrogen isotope behavior and supports future assessments of tritium behavior under neutron irradiation conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155693"},"PeriodicalIF":2.8,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421683","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

In-pile heat conduction model of the dispersion nuclear fuel plate with particle agglomeration. Part II: Predicting the effective thermal conductivity under the in-pile thermal transfer pattern based on a deep neural network

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

Journal of Nuclear Materials

Pub Date : 2025-02-08 DOI: 10.1016/j.jnucmat.2025.155692

Yingxuan Dong , Xicheng Cao , Xingming Peng , Junnan Lv , Qun Li

In-pile dispersion nuclear fuel plate elements incorporate multiple internal heat sources and evolving interaction layers. The effective thermal conductivity of this complex heat conduction structure exhibits multivariate and highly nonlinear characteristics, requiring accurate prediction by combining numerical methods and machine learning techniques. This study developed an automated high-throughput workflow to rapidly simulate massive number of dispersion meat models with diverse microstructures, and further established the machine learning database for predicting the effective thermal conductivity. Influences of internal heat sources within fuel particles, the particle agglomeration and interaction layers were considered simultaneously in modeling. Through utilizing the database constructed by high-throughput computing, the deep neural network was successfully applied to accurately forecast the effective thermal conductivity of the in-pile dispersion meat. Furthermore, comparisons of the predictive performances determined by different distance-based point pattern analysis methods indicated that the radial distribution function (g(r)) served as the most effective approach for characterizing spatial discreteness in predicting the effective thermal conductivity. This work demonstrates that thermal conductivity of ceramic dispersion nuclear fuel plate elements can be enhanced via optimizing the critical microstructural features.

{"title":"In-pile heat conduction model of the dispersion nuclear fuel plate with particle agglomeration. Part II: Predicting the effective thermal conductivity under the in-pile thermal transfer pattern based on a deep neural network","authors":"Yingxuan Dong , Xicheng Cao , Xingming Peng , Junnan Lv , Qun Li","doi":"10.1016/j.jnucmat.2025.155692","DOIUrl":"10.1016/j.jnucmat.2025.155692","url":null,"abstract":"<div><div>In-pile dispersion nuclear fuel plate elements incorporate multiple internal heat sources and evolving interaction layers. The effective thermal conductivity of this complex heat conduction structure exhibits multivariate and highly nonlinear characteristics, requiring accurate prediction by combining numerical methods and machine learning techniques. This study developed an automated high-throughput workflow to rapidly simulate massive number of dispersion meat models with diverse microstructures, and further established the machine learning database for predicting the effective thermal conductivity. Influences of internal heat sources within fuel particles, the particle agglomeration and interaction layers were considered simultaneously in modeling. Through utilizing the database constructed by high-throughput computing, the deep neural network was successfully applied to accurately forecast the effective thermal conductivity of the in-pile dispersion meat. Furthermore, comparisons of the predictive performances determined by different distance-based point pattern analysis methods indicated that the radial distribution function (g(<em>r</em>)) served as the most effective approach for characterizing spatial discreteness in predicting the effective thermal conductivity. This work demonstrates that thermal conductivity of ceramic dispersion nuclear fuel plate elements can be enhanced via optimizing the critical microstructural features.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155692"},"PeriodicalIF":2.8,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143377876","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

High-resolution characterization of cold working effects on the oxidation behavior of alumina-forming austenitic (AFA) steel in lead-bismuth eutectic (LBE) with 10–6 wt.% oxygen at 600 °C

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

Journal of Nuclear Materials

Pub Date : 2025-02-07 DOI: 10.1016/j.jnucmat.2025.155691

Decang Zhang , Xiaoxin Zhang , Xian Zeng , Jun Zhang , Hao Ren , Fanao Meng , Qingzhi Yan

The solution-annealed and cold-worked AFA steels with surface grinding exposed to LBE at 600 °C with 10^–6 wt.% oxygen were investigated using high-resolution characterization techniques. After 500 h exposure, both steels developed protective oxide scales. After 1700 h, solution-annealed steel exhibited dissolution corrosion pits, because Cr and Al in fine grains formed by recrystallization of the surface-grinding subsurface were depleted, forming internal (Cr, Al)₂O₃ nodules that caused the initially developed oxide scales to exfoliate due to increased local stress. Following cold working, increased dislocations and grain boundaries facilitated replenishment of Cr&Al in the subsurface, maintaining the durability of external (Cr, Al)₂O₃ scales, preventing dissolution attack.

引用次数: 0

Investigation of UO2 doped with Fe2O3 sintered under a reducing atmosphere

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

Journal of Nuclear Materials

Pub Date : 2025-02-07 DOI: 10.1016/j.jnucmat.2025.155684

Tashiema L. Ulrich , Tyler L. Spano , Luke Sadergaski , Ashley E. Shields , Eddie Lopez-Honorato , Grant Helmreich , Benjamin T. Manard , Jason Harp , Andrew T. Nelson

Nuclear fuel produced with trace amounts of transition metal additives is of potential interest for introducing intentional signatures for accelerating nuclear forensics. In this work, we investigate the effects on microstructure, grain size, crystal structure, and stoichiometry when trace amounts of Fe in the form of Fe₂O₃ are added to UO₂. Sintering of compacts with different concentrations, from 250 to 3000 ppmw, was performed under a reducing atmosphere at 1773 K. The persistence of the taggant during the fuel fabrication process as well as its impact on grain size, crystal lattice, and first-neighbor chemistry was evaluated using inductively coupled plasma–optical emission spectroscopy, powder X-ray diffraction (pXRD), Raman spectroscopy, scanning electron microscopy–backscatter electron spectroscopy, energy dispersive spectroscopy, and thermogravimetric analysis. We observed that a negligible amount of Fe was lost during sintering. Our results indicate that the feedstock, Fe₂O₃ transforms into Fe and FeO under the test sintering conditions. For all compositions, metallic Fe precipitate was found in grain boundaries as a secondary phase. The potential incorporation of ionic Fe into the UO₂ unit cell was determined by pXRD and Raman spectra.

{"title":"Investigation of UO2 doped with Fe2O3 sintered under a reducing atmosphere","authors":"Tashiema L. Ulrich , Tyler L. Spano , Luke Sadergaski , Ashley E. Shields , Eddie Lopez-Honorato , Grant Helmreich , Benjamin T. Manard , Jason Harp , Andrew T. Nelson","doi":"10.1016/j.jnucmat.2025.155684","DOIUrl":"10.1016/j.jnucmat.2025.155684","url":null,"abstract":"<div><div>Nuclear fuel produced with trace amounts of transition metal additives is of potential interest for introducing intentional signatures for accelerating nuclear forensics. In this work, we investigate the effects on microstructure, grain size, crystal structure, and stoichiometry when trace amounts of Fe in the form of Fe<sub>2</sub>O<sub>3</sub> are added to UO<sub>2</sub>. Sintering of compacts with different concentrations, from 250 to 3000 ppmw, was performed under a reducing atmosphere at 1773 K. The persistence of the taggant during the fuel fabrication process as well as its impact on grain size, crystal lattice, and first-neighbor chemistry was evaluated using inductively coupled plasma–optical emission spectroscopy, powder X-ray diffraction (pXRD), Raman spectroscopy, scanning electron microscopy–backscatter electron spectroscopy, energy dispersive spectroscopy, and thermogravimetric analysis. We observed that a negligible amount of Fe was lost during sintering. Our results indicate that the feedstock, Fe<sub>2</sub>O<sub>3</sub> transforms into Fe and FeO under the test sintering conditions. For all compositions, metallic Fe precipitate was found in grain boundaries as a secondary phase. The potential incorporation of ionic Fe into the UO<sub>2</sub> unit cell was determined by pXRD and Raman spectra.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155684"},"PeriodicalIF":2.8,"publicationDate":"2025-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143394487","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Investigating the synergistic evolution of H and He with irradiation defects in Fe by a new Fe-H-He ternary potential

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

Journal of Nuclear Materials

Pub Date : 2025-02-07 DOI: 10.1016/j.jnucmat.2025.155687

Hao-Xuan Huang , Yu-Hao Li , Tian-Ren Yang , Hong-Bo Zhou , Jin-Liang Li , Hui-Zhi Ma , Yu-Ze Niu , Xiao-Chun Li , Huiqiu Deng , Guang-Hong Lu

It has been demonstrated that the synergistic effect of hydrogen (H) and helium (He) has a significant promoting effect on the nucleation antungstend growth of irradiation defects, leading to severe deterioration of material performance. In order to explore the synergistic evolution behaviors of H and He with irradiation defects in iron (Fe), we have developed a new Fe-H-He ternary potential through systematic fitting and optimization. During the processes, the interaction between H and He with vacancies and interstitial defects is given prominent consideration. Based on the present potential, we further explored the cooperative evolution behaviors of H-He-void complexes. On the one hand, it is clearly observed that both H and He perfectly form a typical core-shell structure in nanovoids with He inside and H outside, which is consistent with experimental observation. The number of H atoms trapped by He-void complexes will decrease with the increasing of He/V ratios, which is due to the repulsion between H and He. On the other hand, the presence of He-void complexes has a significant promoting effect on the formation of H platelets in Fe. The stress field generated by He-void complexes facilitates the aggregation and initial formation of H platelets, and its range expands as H aggregates, ultimately resulting in the formation of H platelets. The high H concentrations are required to form H platelets at high temperature. More importantly, the formation of H platelets can induce a transition in the local Fe lattice from body-centered cubic to face-centered cubic, indicating a substantial impact on the microstructure and properties of Fe. Consequently, the new Fe-H-He potential provides a powerful tool for exploring the synergistic evolution of H-He-defects and evaluating the performance of Fe in fusion devices.

{"title":"Investigating the synergistic evolution of H and He with irradiation defects in Fe by a new Fe-H-He ternary potential","authors":"Hao-Xuan Huang , Yu-Hao Li , Tian-Ren Yang , Hong-Bo Zhou , Jin-Liang Li , Hui-Zhi Ma , Yu-Ze Niu , Xiao-Chun Li , Huiqiu Deng , Guang-Hong Lu","doi":"10.1016/j.jnucmat.2025.155687","DOIUrl":"10.1016/j.jnucmat.2025.155687","url":null,"abstract":"<div><div>It has been demonstrated that the synergistic effect of hydrogen (H) and helium (He) has a significant promoting effect on the nucleation antungstend growth of irradiation defects, leading to severe deterioration of material performance. In order to explore the synergistic evolution behaviors of H and He with irradiation defects in iron (Fe), we have developed a new Fe-H-He ternary potential through systematic fitting and optimization. During the processes, the interaction between H and He with vacancies and interstitial defects is given prominent consideration. Based on the present potential, we further explored the cooperative evolution behaviors of H-He-void complexes. On the one hand, it is clearly observed that both H and He perfectly form a typical core-shell structure in nanovoids with He inside and H outside, which is consistent with experimental observation. The number of H atoms trapped by He-void complexes will decrease with the increasing of He/V ratios, which is due to the repulsion between H and He. On the other hand, the presence of He-void complexes has a significant promoting effect on the formation of H platelets in Fe. The stress field generated by He-void complexes facilitates the aggregation and initial formation of H platelets, and its range expands as H aggregates, ultimately resulting in the formation of H platelets. The high H concentrations are required to form H platelets at high temperature. More importantly, the formation of H platelets can induce a transition in the local Fe lattice from body-centered cubic to face-centered cubic, indicating a substantial impact on the microstructure and properties of Fe. Consequently, the new Fe-H-He potential provides a powerful tool for exploring the synergistic evolution of H-He-defects and evaluating the performance of Fe in fusion devices.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155687"},"PeriodicalIF":2.8,"publicationDate":"2025-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143377875","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Investigation on the SCC susceptibility of a heat-optimized l-PBF 304L stainless steel in simulated primary water

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

Journal of Nuclear Materials

Pub Date : 2025-02-07 DOI: 10.1016/j.jnucmat.2025.155685

Kai Chen , Yuhao Zhou , Fawen Zhu , Shixin Gao , Quanyao Ren , Kun Zhang , Xiaoqing Shang , Lefu Zhang

This study examines the stress corrosion cracking (SCC) behavior of laser powder bed fusion (L-PBF) 304 L stainless steel after solution annealing (SA) in high-temperature water environments. SCC susceptibility was evaluated under both hydrogenated and oxygenated water environments at 325 °C and 15.5 MPa. The results showed that heat-treated l-PBF 304 L exhibited fully recrystallized equiaxed grains, characterized with high-density annealing twins and a uniform microstructure. Stress-strain tests revealed a marked reduction in elongation and mixed intergranular (IG) and transgranular (TG) cracking, indicating significant SCC susceptibility in both environments. Notably, hydrogenated water conditions led to a higher proportion of TG cracking, suggesting enhanced hydrogen embrittlement (HE), while oxygenated conditions favored IG cracking. Cross-sectional analysis highlighted localized strain concentration at crack tips, and precipitates were found to play a critical role in the initiation and propagation of cracks. These findings provide crucial insights into the relationship between microstructure, environmental factors, and SCC behavior in l-PBF 304 L stainless steel.

{"title":"Investigation on the SCC susceptibility of a heat-optimized l-PBF 304L stainless steel in simulated primary water","authors":"Kai Chen , Yuhao Zhou , Fawen Zhu , Shixin Gao , Quanyao Ren , Kun Zhang , Xiaoqing Shang , Lefu Zhang","doi":"10.1016/j.jnucmat.2025.155685","DOIUrl":"10.1016/j.jnucmat.2025.155685","url":null,"abstract":"<div><div>This study examines the stress corrosion cracking (SCC) behavior of laser powder bed fusion (L-PBF) 304 L stainless steel after solution annealing (SA) in high-temperature water environments. SCC susceptibility was evaluated under both hydrogenated and oxygenated water environments at 325 °C and 15.5 MPa. The results showed that heat-treated l-PBF 304 L exhibited fully recrystallized equiaxed grains, characterized with high-density annealing twins and a uniform microstructure. Stress-strain tests revealed a marked reduction in elongation and mixed intergranular (IG) and transgranular (TG) cracking, indicating significant SCC susceptibility in both environments. Notably, hydrogenated water conditions led to a higher proportion of TG cracking, suggesting enhanced hydrogen embrittlement (HE), while oxygenated conditions favored IG cracking. Cross-sectional analysis highlighted localized strain concentration at crack tips, and precipitates were found to play a critical role in the initiation and propagation of cracks. These findings provide crucial insights into the relationship between microstructure, environmental factors, and SCC behavior in l-PBF 304 L stainless steel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155685"},"PeriodicalIF":2.8,"publicationDate":"2025-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143377874","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Study of thermal diffusivity degradation on Cu-OFE copper due to proton and self-ion irradiation using in situ transient grating spectroscopy

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

Journal of Nuclear Materials

Pub Date : 2025-02-05 DOI: 10.1016/j.jnucmat.2025.155674

Emmanouil Trachanas , Angus Wylie , Andrea Bignami , Nikolaos Gazis , Michael P. Short , Katrin Michel , Carl Alwmark , Evangelos Gazis , Georgios Fikioris , Håkan Danared

<div><div>The operation of next-generation particle accelerator facilities with increased beam power parameters and emphasis on performance and reliability signifies the need for experimental studies on material property degradation. Literature is particularly scarce regarding the effects of ionizing radiation damage on cavity components at applicable radiation conditions, with even less extant work studying the effect of simultaneous irradiation on primary materials and their physical properties such as thermal transport. This study presents the impact on thermal diffusivity in oxygen-free electronic Cu-OFE specimens due to proton and self-ion (<span><math><mi>C</mi><msup><mrow><mi>u</mi></mrow><mrow><mn>3</mn><mo>+</mo></mrow></msup></math></span>) irradiation. Copper samples with the exactly same processing route as the bulk material for European Spallation Source (ESS) Radio-Frequency Quadrupole (RFQ) were characterized with scanning electron microscopy (SEM), X-ray diffraction (XRD) and electron backscatter diffraction (EBSD), followed by irradiation at room temperature with protons and self-ions (<span><math><mi>C</mi><msup><mrow><mi>u</mi></mrow><mrow><mn>3</mn><mo>+</mo></mrow></msup></math></span>) with fluences up to 4.25 × 10<sup>17</sup> <span><math><mtext>p</mtext><mo>/</mo><msup><mrow><mtext>cm</mtext></mrow><mrow><mn>2</mn></mrow></msup></math></span> and <span><math><mn>1.10</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mn>15</mn></mrow></msup></math></span> <span><math><mtext>ions</mtext><mo>/</mo><msup><mrow><mtext>cm</mtext></mrow><mrow><mn>2</mn></mrow></msup></math></span>, <span><math><mn>1.69</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mn>15</mn></mrow></msup></math></span> <span><math><mtext>ions</mtext><mo>/</mo><msup><mrow><mtext>cm</mtext></mrow><mrow><mn>2</mn></mrow></msup></math></span> respectively. <em>In situ</em> ion irradiation transient grating spectroscopy (I<sup>3</sup>TGS) was used in order to monitor radiation-induced changes in thermal diffusivity in real time. The results indicate a significant thermal diffusivity drop only after the proton irradiation (10.86%) in contrast to the results obtained for self-ions (<span><math><mi>C</mi><msup><mrow><mi>u</mi></mrow><mrow><mn>3</mn><mo>+</mo></mrow></msup></math></span>) where no significant changes (<span><math><mo>∼</mo><mn>3</mn><mtext>%</mtext></math></span>) are reported. The results are compared with existing literature and correlated with the nature of post-irradiation defect structures based on the particle type. In the case of protons the contribution of hydrogen implantation in stabilization of defects and on the reduction of thermal diffusivity is discussed. <span><math><msup><mrow><mi>I</mi></mrow><mrow><mn>3</mn></mrow></msup><mi>T</mi><mi>G</mi><mi>S</mi></math></span> in combination with concurrent ion irradiation offers a powerful online diagnostic tool for the evaluation of the impact of operational parameters in particle accelerat

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引用次数: 0

Buffering the oxygen activity of uranium dioxide fuels using niobium and molybdenum as redox additives

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

Journal of Nuclear Materials

Pub Date : 2025-02-03 DOI: 10.1016/j.jnucmat.2025.155647

Philippe Garcia , Erwin Douguet-Bronnec , Franck Fournet-Fayard , Florian Le Hello , Xavière Iltis , Nicolas Tarisien , Jacques Fouletier , Marlu Cesar Steil

We demonstrate that it is possible to control the oxygen activity of a uranium oxide fuel by incorporating additives during the manufacturing process that have a strong redox activity. To this end, we have studied the macroscopic electrochemical properties of three polycrystalline uranium dioxide samples, the first containing both molybdenum and molybdenum dioxide, the second containing niobium dioxide and the last containing no additives other than impurities. During experiments, samples were subjected to reducing and subsequently oxidizing atmospheres with respect to the oxygen partial pressure at which the oxidized and reduced forms of the additives are in equilibrium. In situ monitoring of the electrochemical response of the samples reveals that the presence of redox couples involving either niobium or molybdenum significantly modifies the oxygen activity of a uranium dioxide solid. Scanning Electron Microscopy (SEM), Electron Back-Scatter Diffraction (EBSD) and Energy Dispersive X-ray analysis (EDX) prior to and following the high temperature measurements, confirm the in situ data: the final microstructure purports to show that the electrochemical activity is controlled by the Mo/MoO₂ and the NbO₂/Nb₂O₅ couples for the molybdenum and niobium containing samples respectively.

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引用次数: 0

Characterization of plasma in the HiPIMS facility for simulation of fusion-oriented W co-deposition in He/H2 and He/D2 environments

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

Journal of Nuclear Materials

Pub Date : 2025-02-03 DOI: 10.1016/j.jnucmat.2025.155675

Maksim M. Kharkov , Gleb S. Lomonosov , Mikhail S. Novikov , Dobrynya V. Kolodko , Alexander V. Tumarkin , Vitaly S. Efimov , Olga V. Ogorodnikova , Andrey V. Kaziev

A new facility based on a HiPIMS (high-power impulse magnetron sputtering) discharge plasma with a hot tungsten (W) target for the investigation of co-deposition of W with helium (He) and hydrogen (H₂, D₂) isotopes on W substrate was developed. To find the ion fluxes from He/H₂ and He/D₂ plasma together with W ions from W target incident on installed samples, the ion mass spectrometric study of a HiPIMS plasma was performed. The results reveal a mixed composition of ion fluxes consisting mainly of monoatomic H⁺(D⁺) and He⁺, diatomic H₂⁺(D₂⁺) and HeH⁺, and triatomic H₃⁺ ions. The content of tungsten ions does not exceed 2 % in any of the explored discharge regimes. The contributions of different species to the total flux are strongly affected by the discharge pulsing parameters, especially pulsed power density. For pulsed power densities below 200 W/cm², H₃⁺ ions in the He/H₂ or D₃⁺ in He/D₂ gas mixtures dominate in the ion flux. The opposite situation of monoatomic ion fraction (H⁺ and D⁺) prevailing over molecular ions is observed in the case of the pulse power density above the values: 400 W/cm² for prevailing of H⁺ in He/H₂ gas mixture discharge, and 900 W/cm² for prevailing of D⁺ in He/D₂ gas mixture discharge. The most suitable discharge modes are selected for future simulation of W co-deposition processes with He/H₂ and He/D₂ gas mixtures. Preliminary data on the thickness of co-deposits in various regimes have been obtained. The He retention in W irradiated at the HiPIMS facility in hydrogen plasma with ∼ 10 % He ion seeding was validated by comparison with the experimental data after irradiation of W under the same conditions in a linear plasma device.

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引用次数: 0

Effect of irradiation parameters on defect evolution in neutron irradiated tungsten

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

Journal of Nuclear Materials

Pub Date : 2025-02-03 DOI: 10.1016/j.jnucmat.2025.155673

M. Klimenkov , U. Jäntsch , M. Rieth , H.C. Schneider , D. Terentyev , W. Van Renterghem

The article presents a microstructural examination of neutron-irradiated tungsten (W), that was irradiated to four damage doses of 0.1 dpa, 0.2 dpa, 0.5 dpa and 0.8 dpa and at four temperatures of 600°C, 800°C, 900°C and 1200°C in the BR2 material test reactor (Mol, Belgium). The irradiation parameters cover a wide range that enables a comprehensive study of defect formation and evolution. The experimental work includes imaging and quantitative analysis of radiation induced voids and dislocation loops as well as the visualization of the distribution of the transmutation induced Re and Os. It demonstrates the dose- and temperature-dependent evolution of defect's size and number density as well as the segregation behavior of Re and Os at these defects. It was proven that the size of the defects increases, while their number density decreases with increasing damage dose. The formation of nanometer-sized Re–Os precipitates with elongated shape was detected in samples irradiated at 0.8 dpa. The large-scale EDX analysis showed the influence of grain and sub-grain boundaries as well as line dislocations on defect formation and thus on Re and Os-segregation behavior.

{"title":"Effect of irradiation parameters on defect evolution in neutron irradiated tungsten","authors":"M. Klimenkov , U. Jäntsch , M. Rieth , H.C. Schneider , D. Terentyev , W. Van Renterghem","doi":"10.1016/j.jnucmat.2025.155673","DOIUrl":"10.1016/j.jnucmat.2025.155673","url":null,"abstract":"<div><div>The article presents a microstructural examination of neutron-irradiated tungsten (W), that was irradiated to four damage doses of 0.1 dpa, 0.2 dpa, 0.5 dpa and 0.8 dpa and at four temperatures of 600°C, 800°C, 900°C and 1200°C in the BR2 material test reactor (Mol, Belgium). The irradiation parameters cover a wide range that enables a comprehensive study of defect formation and evolution. The experimental work includes imaging and quantitative analysis of radiation induced voids and dislocation loops as well as the visualization of the distribution of the transmutation induced Re and Os. It demonstrates the dose- and temperature-dependent evolution of defect's size and number density as well as the segregation behavior of Re and Os at these defects. It was proven that the size of the defects increases, while their number density decreases with increasing damage dose. The formation of nanometer-sized Re–Os precipitates with elongated shape was detected in samples irradiated at 0.8 dpa. The large-scale EDX analysis showed the influence of grain and sub-grain boundaries as well as line dislocations on defect formation and thus on Re and Os-segregation behavior.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155673"},"PeriodicalIF":2.8,"publicationDate":"2025-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143377657","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0