Journal of Nuclear Materials最新文献

英文中文

Effects of B4C particles on Al-TiO2 reaction in Al matrix composite

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

Journal of Nuclear Materials

Pub Date : 2025-03-23 DOI: 10.1016/j.jnucmat.2025.155776

B.M. Shi , Y. Liu , S.Z. Zhu , Y.N. Zan , Q.Z. Wang , B.L. Xiao , Z.Y. Ma

In this study, composites were produced using powder metallurgy with Al-TiO₂ and Al-TiO₂-B₄C as raw materials, and the impact of B₄C on the reaction between Al and TiO₂ was investigated. Hot-pressing was employed at temperatures of 580 °C, 600 °C, 620 °C, and 640 °C to fabricate the composites, followed by analysis of their microstructure and mechanical properties. The results revealed that at a temperature of 580 °C, Al₃Ti and Al₂O₃ were formed in the Al-TiO₂ composite system, and when the hot-pressing temperature reached 620 °C, all TiO₂ within the matrix was completely consumed in the reaction with Al. In contrast, for the Al-TiO₂-B₄C composite system, B₄C exhibited a noticeable inhibiting effect on the reaction. At a hot-pressing temperature of 580 °C, no discernible reaction occurred between TiO₂ and the matrix; while at temperatures of 600 °C, 620 °C, and 640 °C, only a small amount of Al₃Ti and Al₂O₃ appeared in the matrix. Consequently, the strength of composites fabricated by Al-TiO₂ increased with increasing hot-pressing temperature; whereas for composites made from Al-TiO₂-B₄C hot-pressed at different temperatures showed similar strengths.

{"title":"Effects of B4C particles on Al-TiO2 reaction in Al matrix composite","authors":"B.M. Shi , Y. Liu , S.Z. Zhu , Y.N. Zan , Q.Z. Wang , B.L. Xiao , Z.Y. Ma","doi":"10.1016/j.jnucmat.2025.155776","DOIUrl":"10.1016/j.jnucmat.2025.155776","url":null,"abstract":"<div><div>In this study, composites were produced using powder metallurgy with Al-TiO<sub>2</sub> and Al-TiO<sub>2</sub>-B<sub>4</sub>C as raw materials, and the impact of B<sub>4</sub>C on the reaction between Al and TiO<sub>2</sub> was investigated. Hot-pressing was employed at temperatures of 580 °C, 600 °C, 620 °C, and 640 °C to fabricate the composites, followed by analysis of their microstructure and mechanical properties. The results revealed that at a temperature of 580 °C, Al<sub>3</sub>Ti and Al<sub>2</sub>O<sub>3</sub> were formed in the Al-TiO<sub>2</sub> composite system, and when the hot-pressing temperature reached 620 °C, all TiO<sub>2</sub> within the matrix was completely consumed in the reaction with Al. In contrast, for the Al-TiO<sub>2</sub>-B<sub>4</sub>C composite system, B<sub>4</sub>C exhibited a noticeable inhibiting effect on the reaction. At a hot-pressing temperature of 580 °C, no discernible reaction occurred between TiO<sub>2</sub> and the matrix; while at temperatures of 600 °C, 620 °C, and 640 °C, only a small amount of Al<sub>3</sub>Ti and Al<sub>2</sub>O<sub>3</sub> appeared in the matrix. Consequently, the strength of composites fabricated by Al-TiO<sub>2</sub> increased with increasing hot-pressing temperature; whereas for composites made from Al-TiO<sub>2</sub>-B<sub>4</sub>C hot-pressed at different temperatures showed similar strengths.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155776"},"PeriodicalIF":2.8,"publicationDate":"2025-03-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143725314","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

O2- affected corrosion of nickel-based alloy in LiCl-KCl molten salt for pyroprocessing of spent nuclear fuel

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

Journal of Nuclear Materials

Pub Date : 2025-03-22 DOI: 10.1016/j.jnucmat.2025.155773

Chengyu Wang, Xiaorui Xu, Yang Tong, Wentao Zhou, Yafei Wang

Pyroprocessing of spent nuclear fuel is an important step in nuclear fuel cycle with the purpose of minimizing the radiation impact on environment and recycling useful elements. One major concern in the development of pyroprocessing is the structural material corrosion in LiCl-KCl molten salt electrolyte, which is the core component of the spent nuclear fuel reprocessing. To understand the influence of O^2- impurity on the structural material corrosion in LiCl-KCl molten salt, the study investigates the corrosion behavior of Inconel 600 alloy in LiCl-KCl molten salt containing different amounts of O^2- impurities at 500 °C by immersion corrosion experiments and electrochemical analyses.The results indicate that O^2- could accelerate the high-temperature corrosion behavior. The alloy surface developed a bilayer oxide structure following 500-hour exposure to molten salt with 3 mol % O^2-, identified as a 3.36-µm-thick outer Li₂NiO₂ layer and a 1.11-µm-thick inner LiCrO₂ layer. Based on the corrosion behaviors of Inconel 600 alloy in LiCl-KCl molten salt with different amounts of O^2- impurities, a new O^2- affected corrosion mechanism involving the formation of dual-layer oxide films and nickel-rich particles was proposed, which could provide new insights for the corrosion study in the applications of pyroprocessing of spent nuclear fuel.

{"title":"O2- affected corrosion of nickel-based alloy in LiCl-KCl molten salt for pyroprocessing of spent nuclear fuel","authors":"Chengyu Wang, Xiaorui Xu, Yang Tong, Wentao Zhou, Yafei Wang","doi":"10.1016/j.jnucmat.2025.155773","DOIUrl":"10.1016/j.jnucmat.2025.155773","url":null,"abstract":"<div><div>Pyroprocessing of spent nuclear fuel is an important step in nuclear fuel cycle with the purpose of minimizing the radiation impact on environment and recycling useful elements. One major concern in the development of pyroprocessing is the structural material corrosion in LiCl-KCl molten salt electrolyte, which is the core component of the spent nuclear fuel reprocessing. To understand the influence of O<sup>2-</sup> impurity on the structural material corrosion in LiCl-KCl molten salt, the study investigates the corrosion behavior of Inconel 600 alloy in LiCl-KCl molten salt containing different amounts of O<sup>2-</sup> impurities at 500 °C by immersion corrosion experiments and electrochemical analyses.The results indicate that O<sup>2-</sup> could accelerate the high-temperature corrosion behavior. The alloy surface developed a bilayer oxide structure following 500-hour exposure to molten salt with 3 mol % O<sup>2-</sup>, identified as a 3.36-µm-thick outer Li<sub>2</sub>NiO<sub>2</sub> layer and a 1.11-µm-thick inner LiCrO<sub>2</sub> layer. Based on the corrosion behaviors of Inconel 600 alloy in LiCl-KCl molten salt with different amounts of O<sup>2-</sup> impurities, a new O<sup>2-</sup> affected corrosion mechanism involving the formation of dual-layer oxide films and nickel-rich particles was proposed, which could provide new insights for the corrosion study in the applications of pyroprocessing of spent nuclear fuel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155773"},"PeriodicalIF":2.8,"publicationDate":"2025-03-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143807811","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

Impact of composition and cold work on SCC of high Si-bearing stainless steel simulating radiation-induced segregation under simulated PWR primary water

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

Journal of Nuclear Materials

Pub Date : 2025-03-22 DOI: 10.1016/j.jnucmat.2025.155775

Caibo Xie , Deng Pan , Hao Liu , Zhanpeng Lu , Tongming Cui , Junjie Chen , Xinhe Xu

Thick block alloys simulating radiation-induced segregation effects of compositions at grain boundaries were designed and prepared, and deformation due to irradiation hardening was simulated by cold working treatments. The stress corrosion cracking (SCC) behavior of 17Cr-13Ni-3Si, 13Cr-17Ni, and 13Cr-17Ni-3Si alloys in the as-cast and cold-worked states was studied by crack growth rate (CGR) tests using compact tension specimens in hydrogenated high-temperature water. The fracture surfaces of the three as-cast state simulated alloys showed local SCC indications, while the fracture surfaces of all cold-worked state simulated alloys showed typical intergranular SCC characteristics. In the absence of cold work, the single Si addition has no noticeable effect on SCC propagation. Stress-accelerated grain boundary oxidation acts as the initiation site for intergranular cracks. Cold-worked low Cr alloy with high Si content significantly promotes SCC growth in hydrogenated water. The effect of alloy composition and strain hardening on SCC and its implications for irradiation-assisted SCC were discussed.

{"title":"Impact of composition and cold work on SCC of high Si-bearing stainless steel simulating radiation-induced segregation under simulated PWR primary water","authors":"Caibo Xie , Deng Pan , Hao Liu , Zhanpeng Lu , Tongming Cui , Junjie Chen , Xinhe Xu","doi":"10.1016/j.jnucmat.2025.155775","DOIUrl":"10.1016/j.jnucmat.2025.155775","url":null,"abstract":"<div><div>Thick block alloys simulating radiation-induced segregation effects of compositions at grain boundaries were designed and prepared, and deformation due to irradiation hardening was simulated by cold working treatments. The stress corrosion cracking (SCC) behavior of 17Cr-13Ni-3Si, 13Cr-17Ni, and 13Cr-17Ni-3Si alloys in the as-cast and cold-worked states was studied by crack growth rate (CGR) tests using compact tension specimens in hydrogenated high-temperature water. The fracture surfaces of the three as-cast state simulated alloys showed local SCC indications, while the fracture surfaces of all cold-worked state simulated alloys showed typical intergranular SCC characteristics. In the absence of cold work, the single Si addition has no noticeable effect on SCC propagation. Stress-accelerated grain boundary oxidation acts as the initiation site for intergranular cracks. Cold-worked low Cr alloy with high Si content significantly promotes SCC growth in hydrogenated water. The effect of alloy composition and strain hardening on SCC and its implications for irradiation-assisted SCC were discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155775"},"PeriodicalIF":2.8,"publicationDate":"2025-03-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143715458","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

Computational analysis of plasma-wall interactions in beryllium: A detailed study of physical and chemically assisted physical sputtering

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

Journal of Nuclear Materials

Pub Date : 2025-03-20 DOI: 10.1016/j.jnucmat.2025.155758

Nima Fakhrayi Mofrad , Juri Romazanov , Roy Schumacher , Andrea E. Sand

Understanding plasma-wall interactions is one of the main challenges in the design and development of fusion reactors. Among the primary effects of these interactions is the erosion of plasma-facing components through physical or chemical sputtering, which can limit the availability and performance of the device. We simulate this phenomenon in beryllium surfaces with varying concentrations of hydrogen isotopes using atomistic molecular dynamics. Special attention is given to chemical sputtering and the overall behavior of molecules emitted from the surface. Our findings indicate that the balance between physical and chemical sputtering is considerably affected by isotope type, impact energy, and incident angle of the plasma particle. We compare the results with predictions from SDTrimSP, a tool that utilizes the more computationally efficient binary collision approximation, to elucidate the conditions where the higher accuracy of molecular dynamics is needed. Moreover, we highlight the effect of surface temperature, which determines the concentration of hydrogen isotopes in the surface layers, on the contribution of chemical sputtering to total erosion, and the types of sputtered molecules. Lastly, we demonstrate that the escape energies and angles of the sputtered species are also significantly influenced by the impact energy and angle of the plasma particles.

{"title":"Computational analysis of plasma-wall interactions in beryllium: A detailed study of physical and chemically assisted physical sputtering","authors":"Nima Fakhrayi Mofrad , Juri Romazanov , Roy Schumacher , Andrea E. Sand","doi":"10.1016/j.jnucmat.2025.155758","DOIUrl":"10.1016/j.jnucmat.2025.155758","url":null,"abstract":"<div><div>Understanding plasma-wall interactions is one of the main challenges in the design and development of fusion reactors. Among the primary effects of these interactions is the erosion of plasma-facing components through physical or chemical sputtering, which can limit the availability and performance of the device. We simulate this phenomenon in beryllium surfaces with varying concentrations of hydrogen isotopes using atomistic molecular dynamics. Special attention is given to chemical sputtering and the overall behavior of molecules emitted from the surface. Our findings indicate that the balance between physical and chemical sputtering is considerably affected by isotope type, impact energy, and incident angle of the plasma particle. We compare the results with predictions from SDTrimSP, a tool that utilizes the more computationally efficient binary collision approximation, to elucidate the conditions where the higher accuracy of molecular dynamics is needed. Moreover, we highlight the effect of surface temperature, which determines the concentration of hydrogen isotopes in the surface layers, on the contribution of chemical sputtering to total erosion, and the types of sputtered molecules. Lastly, we demonstrate that the escape energies and angles of the sputtered species are also significantly influenced by the impact energy and angle of the plasma particles.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"609 ","pages":"Article 155758"},"PeriodicalIF":2.8,"publicationDate":"2025-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682846","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

First-principles calculation of oxygen absorption, solution and diffusion in Zr and its coatings (Cr, Cr/α-Al2O3)

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

Journal of Nuclear Materials

Pub Date : 2025-03-19 DOI: 10.1016/j.jnucmat.2025.155759

Yiying Yang , Xiaokai Zhou , Rongchuan Li , Tianyuan Xin , Jiaxuan Si , Yaojun Li , Yuexia Wang

This work used first-principles calculations to investigate the O behavior near the Zr(0001), Cr(100), and α-Al₂O₃(0001) surfaces, as well as in the Zr(0001)/Cr(100) and Cr(100)/α-Al₂O₃(0001) interfaces. The calculation of absorption, solution, and migration energies, and the analysis of interlayer distances, electronic properties, and jump rates, demonstrated that the α-Al₂O₃(0001) surface possesses the highest resistance to O ingression among these three surfaces. Moreover, the evaluation of interfacial bonding strength revealed that Cr layers are competent to tightly adhere to both Zr substrate and α-Al₂O₃ coating. Consequently, this work proposed a promising coating, Cr/α-Al₂O₃ composite coating with high adhesion work for resisting the delamination of the composite coating. The composite coating exhibits better oxidation resistance than both Zr substrate and Cr coating, which primarily owes to the outer α-Al₂O₃ coating as a shielding layer since O atoms can easily cross the interfaces according to the calculation of migration energies. This work provided theoretical guidance for the development of zircaloy cladding coating materials.

{"title":"First-principles calculation of oxygen absorption, solution and diffusion in Zr and its coatings (Cr, Cr/α-Al2O3)","authors":"Yiying Yang , Xiaokai Zhou , Rongchuan Li , Tianyuan Xin , Jiaxuan Si , Yaojun Li , Yuexia Wang","doi":"10.1016/j.jnucmat.2025.155759","DOIUrl":"10.1016/j.jnucmat.2025.155759","url":null,"abstract":"<div><div>This work used first-principles calculations to investigate the O behavior near the Zr(0001), Cr(100), and <em>α</em>-Al<sub>2</sub>O<sub>3</sub>(0001) surfaces, as well as in the Zr(0001)/Cr(100) and Cr(100)/<em>α</em>-Al<sub>2</sub>O<sub>3</sub>(0001) interfaces. The calculation of absorption, solution, and migration energies, and the analysis of interlayer distances, electronic properties, and jump rates, demonstrated that the <em>α</em>-Al<sub>2</sub>O<sub>3</sub>(0001) surface possesses the highest resistance to O ingression among these three surfaces. Moreover, the evaluation of interfacial bonding strength revealed that Cr layers are competent to tightly adhere to both Zr substrate and <em>α</em>-Al<sub>2</sub>O<sub>3</sub> coating. Consequently, this work proposed a promising coating, Cr/<em>α</em>-Al<sub>2</sub>O<sub>3</sub> composite coating with high adhesion work for resisting the delamination of the composite coating. The composite coating exhibits better oxidation resistance than both Zr substrate and Cr coating, which primarily owes to the outer <em>α</em>-Al<sub>2</sub>O<sub>3</sub> coating as a shielding layer since O atoms can easily cross the interfaces according to the calculation of migration energies. This work provided theoretical guidance for the development of zircaloy cladding coating materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"609 ","pages":"Article 155759"},"PeriodicalIF":2.8,"publicationDate":"2025-03-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143695922","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

Electrochemical behavior and extraction of Nd (III) on reactive zinc electrode in LiCl-KCl melts

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

Journal of Nuclear Materials

Pub Date : 2025-03-18 DOI: 10.1016/j.jnucmat.2025.155751

De-Bin Ji , Rui-Long Liang , Heng-Yu Tan , Shao-Ting Zhang , De-Qiang Ji , Hong-Jun Wu

The efficient extraction of neodymium is of great significance in the dry reprocessing process for the nuclear fuel cycle. In this research, the extraction of Nd in a LiCl-KCl molten salt system was achieved by potentiostatic electrolysis and galvanostatic electrolysis using Zn as the active cathode. A variety of electrochemical methods were employed to deeply investigate the electrochemical behavior of Nd on an inert Mo electrode, galvanized (Zn-coated) Mo electrode, and liquid Zn electrode in the LiCl-KCl melt. The six Zn-Nd intermetallic compounds formed on the Zn-coated Mo electrode were analyzed and determined in detail, and the thermodynamic properties of these six Zn-Nd intermetallic compounds were investigated. Furthermore, the activity coefficients and solubilities of Nd on the liquid Zn electrode were calculated at different temperatures. This study provides some fundamental data for spent fuel reprocessing technology and is expected to help advance the sustainable development of nuclear energy.

引用次数: 0

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

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

Journal of Nuclear Materials

Pub Date : 2025-03-18 DOI: 10.1016/j.jnucmat.2025.155746

William A. Hanson, Daniele Salvato, Adam B. Robinson, Nancy J. Lybeck, Jan-Fong Jue, Tammy L. Trowbridge, Jatuporn Burns, Fidelma G. Di Lemma, Charlyne A. Smith, Margaret A. Marshall, Dennis D. Keiser Jr., Jeffrey J. Giglio, James I. Cole

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. The swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

{"title":"Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates","authors":"William A. Hanson, Daniele Salvato, Adam B. Robinson, Nancy J. Lybeck, Jan-Fong Jue, Tammy L. Trowbridge, Jatuporn Burns, Fidelma G. Di Lemma, Charlyne A. Smith, Margaret A. Marshall, Dennis D. Keiser Jr., Jeffrey J. Giglio, James I. Cole","doi":"10.1016/j.jnucmat.2025.155746","DOIUrl":"10.1016/j.jnucmat.2025.155746","url":null,"abstract":"<div><div>Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. The swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"609 ","pages":"Article 155746"},"PeriodicalIF":2.8,"publicationDate":"2025-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143643135","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

Identification of uranium oxidation states using oxygen K-edge scanning transmission X-ray microscopy

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

Journal of Nuclear Materials

Pub Date : 2025-03-18 DOI: 10.1016/j.jnucmat.2025.155754

Rachel E. Lim , Alexander A. Baker , Alexander S. Ditter , S. Olivia Gunther , David K. Shuh , Jack M. Mayer , Matthew A. Marcus , Scott B. Donald , Brandon W. Chung

The field of nuclear forensics is growing in importance, and the increasing capabilities at synchrotron radiation light sources enable non-destructive characterization of oxide particles with better spatial, compositional, and oxidation state speciation resolution than ever before. Uranium oxide particles derived from multiple wet chemical processing methods were examined using a scanning transmission X-ray microscope (STXM), and a weakly-supervised method was developed to automatically analyze the collected data. Multiple uranium oxidation states were observed and quantified within and between samples, yielding information about differences between particles produced via the various processing routes.

引用次数: 0

Semi-integral LOCA testing of Cr-coated Optimized ZIRLOTM claddings

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

Journal of Nuclear Materials

Pub Date : 2025-03-18 DOI: 10.1016/j.jnucmat.2025.155766

Ioannis Alakiozidis , Marc Lopes Nunes de Sousa , Axel Gauthier , Callum Hunt , Mia Maric , Antoine Ambard , Zaheen Shah , Philipp Frankel

Chromium (Cr)-coatings on zirconium-(Zr) based claddings have emerged as a promising short-term solution to enhance the accident tolerance of fuel assemblies in pressurised water reactors (PWRs) during loss-of-coolant accidents (LOCAs). In this study, we tested a large number (36 rods in total, each 30cm long) of uncoated and Cr-coated Optimized ZIRLO^TM claddings under thermomechanical conditions that closely resemble a real LOCA. A unique experimental apparatus was employed to integrate multiple LOCA effects into a single test sequence, enabling a more accurate prediction of the performance of Cr-coatings and degradation mechanisms of the coated claddings. More specifically, the test sequence included: i) thermal ramping from 350–1200°C under varying internal pressures and heating rates in flowing steam; ii) isothermal steam oxidation at 1200°C for different durations; ii) cooling to 700°C followed by water quenching to 135°C; iv) partial-axial constraint at 135°C with load hold of 540N for 20s. Various characterisation techniques, including optical and scanning electron microscopy (SEM), 3D laser scanning, electron backscattered diffraction (EBSD), hardness testing and hydrogen analysis, were used to characterise the post-LOCA cladding microstructures. We found that Cr-coatings increased the burst temperature of uncoated claddings by ∼ 25–150°C and reduced the strain-to-burst and cladding deformation within 20 mm away from the burst opening. The magnitude of these improvements depended on the initial testing conditions and were more pronounced for the helium-propelled cold spray (HCS) coating, while less pronounced for the nitrogen-propelled CS (NCS) and physical vapour deposition (PVD) coatings. Additionally, we found that Cr-coatings increased the time threshold before significant cladding embrittlement by ∼100–555s compared to uncoated claddings. Finally, we concluded that when multiple LOCA effects are considered, predictions of additional coping time during a LOCA provided by the Cr-coatings are more conservative compared to single-factor tests.

锆（Zr）基覆层上的铬（Cr）涂层已成为一种很有前途的短期解决方案，可提高压水堆（PWR）燃料组件在失冷事故（LOCA）期间的事故耐受性。在这项研究中，我们在与真实 LOCA 非常相似的热机械条件下测试了大量（共 36 根棒，每根 30 厘米长）无涂层和有铬涂层的优化 ZIRLOTM 包壳。采用了一种独特的实验装置，将多种 LOCA 效应整合到一个测试序列中，从而能够更准确地预测铬涂层的性能和涂层包层的降解机制。更具体地说，测试序列包括：i) 在流动蒸汽中以不同的内部压力和加热速率从 350°C 升温到 1200°C；ii) 在 1200°C 进行不同持续时间的等温蒸汽氧化；ii) 冷却到 700°C，然后水淬到 135°C；iv) 在 135°C 进行部分轴向约束，负载保持 540N 20 秒。我们采用了各种表征技术，包括光学和扫描电子显微镜 (SEM)、三维激光扫描、电子背散射衍射 (EBSD)、硬度测试和氢分析，来表征 LOCA 后包层的微观结构。我们发现，Cr 涂层将未涂层包层的爆裂温度提高了 25-150°C 左右，并降低了爆裂应变和距爆裂口 20 毫米范围内的包层变形。这些改善的幅度取决于初始测试条件，氦气推进冷喷（HCS）涂层的改善幅度更大，而氮气推进 CS（NCS）和物理气相沉积（PVD）涂层的改善幅度较小。此外，我们还发现，与未涂层的覆层相比，Cr 涂层将覆层发生明显脆化之前的时间阈值提高了 100-555 秒。最后，我们得出结论，当考虑到多重 LOCA 影响时，与单因素测试相比，Cr 涂层提供的 LOCA 期间额外应对时间的预测更为保守。

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引用次数: 0

Low temperature neutron irradiation stability of Zirconium hydride and Yttrium hydride

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

Journal of Nuclear Materials

Pub Date : 2025-03-18 DOI: 10.1016/j.jnucmat.2025.155770

D.J. Sprouster , M. Ouyang , N. Cetiner , P. Negi , A. Sharma , D. Bhardwaj , Y. Huang , X. Hu , K. Shirvan , L.L. Snead

Metal hydrides, including ZrH_x and YH_x, are of particular interest for advanced thermal fission reactors as they have high neutron moderating power and can be used at relatively high temperatures. They have direct applications as core components including as a moderating addition in nuclear fuel, and as neutron reflectors or moderators. Understanding their thermal and irradiation-induced property changes are important to their engineering application. Specifically, evolving metal hydrogen ratios are of critical importance. In this work we discuss the post-irradiation examination of neutron irradiated ZrH_2-x and YH_2-x specimens. We employ multiple characterization techniques including X-ray diffraction, scanning electron microscopy and thermophysical (thermal diffusivity) to determine the irradiation-induced macro- and microstructural evolution as a function of irradiation temperature. We readily quantify degradations in the thermal diffusivity, changes in lattice parameters, and an increase in metallic Zr indicative of hydrogen release in ZrH_2-x specimens. Interestingly, minimal-to-nil change in the metallic Y fraction was quantifiable in the YH_2-x specimens and modest changes in the thermal diffusivity occur for the temperature and dose studied. The loss of hydrogen in the ZrH_2-x samples is related to an apparent irradiation-accelerated desorption of hydrogen by the high ionizing radiation components (gamma, epithermal and fast neutron fluxes) from the in-core neutron irradiation. The most apparent feature from the microstructural analysis for both metal hydrides was a temperature-dependent decrease in the X-ray diffraction peak broadening, attributable to changes in the number and makeup of the two-dimensional defects. These results and trends improve both the fundamental understanding of neutron-solid interactions, and the development of such an important class of core materials.

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